Sample records for srs high-level waste

In 1997, the first two United States Department of Energy (US DOE) highlevelwaste tanks (Tanks 17-F and 20-F: Type IV, single shell tanks) were taken out of service (permanently closed) at the Savannah River Site (SRS). In 2012, the DOE plans to remove from service two additional Savannah River Site (SRS) Type IV high-levelwaste tanks, Tanks 18-F and 19-F. These tanks were constructed in the late 1950's and received low-heat waste and do not contain cooling coils. Operational closure of Tanks 18-F and 19-F is intended to be consistent with the applicable requirements of the Resource Conservation and Recovery Act (RCRA) and the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) and will be performed in accordance with South Carolina Department of Health and Environmental Control (SCDHEC). The closure will physically stabilize two 4.92E+04 cubic meter (1.3 E+06 gallon) carbon steel tanks and isolate and stabilize any residual contaminants left in the tanks. The closure will also fill, physically stabilize and isolate ancillary equipment abandoned in the tanks. A Performance Assessment (PA) has been developed to assess the long-term fate and transport of residual contamination in the environment resulting from the operational closure of the F-Area Tank Farm (FTF) waste tanks. Next generation flowable, zero-bleed cementitious grouts were designed, tested, and specified for closing Tanks 18-F and 19-F and for filling the abandoned equipment. Fill requirements were developed for both the tank and equipment grouts. All grout formulations were required to be alkaline with a pH of 12.4 and chemically reduction potential (Eh) of -200 to -400 to stabilize selected potential contaminants of concern. This was achieved by including Portland cement and Grade 100 slag in the mixes, respectively. Ingredients and proportions of cementitious reagents were selected and adjusted, respectively, to support the mass placement strategy developed by closure

In 1997, the first two United States Department of Energy (US DOE) highlevelwaste tanks (Tanks 17-F and 20-F: Type IV, single shell tanks) were taken out of service (permanently closed) at the Savannah River Site (SRS). In 2012, the DOE plans to remove from service two additional Savannah River Site (SRS) Type IV high-levelwaste tanks, Tanks 18-F and 19-F. These tanks were constructed in the late 1950's and received low-heat waste and do not contain cooling coils. Operational closure of Tanks 18-F and 19-F is intended to be consistent with the applicable requirements of the Resource Conservation and Recovery Act (RCRA) and the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) and will be performed in accordance with South Carolina Department of Health and Environmental Control (SCDHEC). The closure will physically stabilize two 4.92E+04 cubic meter (1.3 E+06 gallon) carbon steel tanks and isolate and stabilize any residual contaminants left in the tanks. Ancillary equipment abandoned in the tanks will also be filled to the extent practical. A Performance Assessment (PA) has been developed to assess the long-term fate and transport of residual contamination in the environment resulting from the operational closure of the F-Area Tank Farm (FTF) waste tanks. Next generation flowable, zero-bleed cementitious grouts were designed, tested, and specified for closing Tanks 18-F and 19-F and for filling the abandoned equipment. Fill requirements were developed for both the tank and equipment grouts. All grout formulations were required to be alkaline with a pH of 12.4 and to be chemically reducing with a reduction potential (Eh) of -200 to -400. Grouts with this chemistry stabilize potential contaminants of concern. This was achieved by including Portland cement and Grade 100 slag in the mixes, respectively. Ingredients and proportions of cementitious reagents were selected and adjusted to support the mass placement strategy developed by

The process of recovering the waste in storage tanks at the Savannah River Site (SRS) typically requires mixing the contents of the tank to ensure uniformity of the discharge stream. Mixing is accomplished with one to four slurry pumps located within the tank liquid. The slurry pump may be fixed in position or they may rotate depending on the specific mixing requirements. The high-levelwaste in Tank 48 contains insoluble solids in the form of potassium tetraphenyl borate compounds (KTPB), monosodium titanate (MST), and sludge. Tank 48 is equipped with 4 slurry pumps, which are intended to suspend the insoluble solids prior to transfer of the waste to the Fluidized Bed Steam Reformer (FBSR) process. The FBSR process is being designed for a normal feed of 3.05 wt% insoluble solids. A chemical characterization study has shown the insoluble solids concentration is approximately 3.05 wt% when well-mixed. The project is requesting a Computational Fluid Dynamics (CFD) mixing study from SRNL to determine the solids behavior with 2, 3, and 4 slurry pumps in operation and an estimate of the insoluble solids concentration at the suction of the transfer pump to the FBSR process. The impact of cooling coils is not considered in the current work. The work consists of two principal objectives by taking a CFD approach: (1) To estimate insoluble solids concentration transferred from Tank 48 to the Waste Feed Tank in the FBSR process and (2) To assess the impact of different combinations of four slurry pumps on insoluble solids suspension and mixing in Tank 48. For this work, several different combinations of a maximum of four pumps are considered to determine the resulting flow patterns and local flow velocities which are thought to be associated with sludge particle mixing. Two different elevations of pump nozzles are used for an assessment of the flow patterns on the tank mixing. Pump design and operating parameters used for the analysis are summarized in Table 1. The baseline

The Defense Waste Processing Facility (DWPF) Savannah River Site is currently immobilizing highlevel nuclear waste sludge by vitrification in borosilicate glass. The processing strategy involves blending a large batch of sludge into a feed tank, washing the sludge to reduce the amount of soluble species, then processing the large ''sludge batch'' through the DWPF. Each sludge batch is tested by the Savannah River National Laboratory (SRNL) using simulants and tests with samples of the radioactive waste to ''qualify'' the batch prior to processing in the DWPF. The DWPF pretreats the sludge by first acidifying the sludge with nitric and formic acid. The ratio of nitric to formic acid is adjusted as required to target a final glass composition that is slightly reducing (the target is for {approx}20% of the iron to have a valence of two in the glass). The formic acid reduces the mercury in the feed to elemental mercury which is steam stripped from the feed. After a concentration step, the glass former (glass frit) is added as a 50 wt% slurry and the batch is concentrated to approximately 50 wt% solids. The feed slurry is then fed to a joule heated melter maintained at 1150 C. The glass must meet both processing (e.g., viscosity and liquidus temperature) and product performance (e.g., durability) constraints The alkali content of the final waste glass is a critical parameter that affects key glass properties (such as durability) as well as the processing characteristics of the waste sludge during the pretreatment and vitrification processes. Increasing the alkali content of the glass has been shown to improve the production rate of the DWPF, but the total alkali in the final glass is limited by constraints on glass durability and viscosity. Two sources of alkali contribute to the final alkali content of the glass: sodium salts in the waste supernate and sodium and lithium oxides in the glass frit added during pretreatment processes. Sodium salts in the waste supernate

In the highlevelwaste tanks at the Savannah River Site (SRS), hydrogen is produced continuously by interaction of the radiation in the tank with water in the waste. Consequently, the vapor spaces of the tanks are purged to prevent the accumulation of H{sub 2} and possible formation of a flammable mixture in a tank. Personnel at SRS have developed an empirical model to predict the rate of H{sub 2} formation in a tank. The basis of this model is the prediction of the G value for H{sub 2} production. This G value is the number of H{sub 2} molecules produced per 100 eV of radiolytic energy absorbed by the waste. Based on experimental studies it was found that the G value for H{sub 2} production from beta radiation and from gamma radiation were essentially equal. The G value for H{sub 2} production from alpha radiation was somewhat higher. Thus, the model has two equations, one for beta/gamma radiation and one for alpha radiation. Experimental studies have also indicated that both G values are decreased by the presence of nitrate and nitrite ions in the waste. These are the main scavengers for the precursors of H{sub 2} in the waste; thus the equations that were developed predict G values for hydrogen production as a function of the concentrations of these two ions in waste. Knowing the beta/gamma and alpha heat loads in the waste allows one to predict the total generation rate for hydrogen in a tank. With this prediction a ventilation rate can be established for each tank to ensure that a flammable mixture is not formed in the vapor space in a tank. Recently personnel at Hanford have developed a slightly different model for predicting hydrogen G values. Their model includes the same precursor for H{sub 2} as the SRS model but also includes an additional precursor not in the SRS model. Including the second precursor for H{sub 2} leads to different empirical equations for predicting the G values for H{sub 2} as a function of the nitrate and nitrite concentrations in

Citizen involvement in DOE's decision-making for the environmental cleanup from DOE's management of its nuclear wastes across the DOE complex has had a positive effect on the cleanup of its SRS site, characterized by an acceleration of cleanup not only for the Transuranic wastes at SRS, but also for DOE's first two closures of HLW tanks, both of which occurred at SRS. The Citizens around SRS had pushed successfully for the closures of Tanks 17 and 20 in 1997, becoming the first closures of HLW tanks under regulatory guidance in the USA. However, since then, HLW tank closures ceased due to a lawsuit, the application of new tank clean-up technology, interagency squabbling between DOE and NRC over tank closure criteria, and finally and almost fatally, from budget pressures. Despite an agreement with its regulators for the closure of Tanks 18 and 19 by the end of calendar year 2012, the outlook in Fall 2011 to close these two tanks had dimmed. It was at this point that the citizens around SRS became reengaged with tank closures, helping DOE to reach its agreed upon milestone. (authors)

The DOE Division of Waste Products through a lead office at Savannah River is developing a program to immobilize all US high-level nuclear waste for terminal disposal. DOE high-levelwastes include those at the Hanford Plant, the Idaho Chemical Processing Plant, and the Savannah River Plant. Commercial high-levelwastes, for which DOE is also developing immobilization technology, include those at the Nuclear Fuel Services Plant and any future commercial fuels reprocessing plants. The first immobilization plant is to be the Defense Waste Processing Facility at Savannah River, scheduled for 1983 project submission to Congress and 1989 operation. Waste forms are still being selected for this plant. Borosilicate glass is currently the reference form, but alternate candidates include concretes, calcines, other glasses, ceramics, and matrix forms.

Closure of the remaining tanks and final disposition of the radioactive waste is a high priority task at both Savannah River Site (SRS) and Hanford. The radioactive waste in the tanks are generally found in layers: supernate (on top) containing soluble fission products, and salt-cake and sludge (on the bottom of the tank) containing insoluble actinides. One strategy for minimizing the waste volume is to segregate the low curie salt waste from the high curie salt supernate by draining the supernate and interstitial salt solution from the salt-cake. The retrieval of the interstitial fluid will require knowledge of relevant properties of salt-cake waste including drainage parameters, more specifically, its hydraulic properties. The hydraulic parameters of the salt-cake have significance with respect to: 1) Kinetics of the retrieval process; and 2) Equilibrium conditions of the drainage. While the saturated hydraulic properties of the salt waste (hydraulic conductivity in the vertical and horizontal direction) can be used to determine the kinetics of the flow through the salt waste, the unsaturated properties are needed in order to assess not only the time frame of tank drainage but also the equilibrium conditions. How much and how fast fluid can be drained at given initial and boundary conditions (atmospheric pressure and temperature) can be analyzed. A series of dissolution and drainage experiments was conducted using S-112, S-109 and Tank 41 simulants in a pilot-scale column (1' diameter, 10' high). The major goal of these experiments was to determine the hydraulic parameters of flow through the column and the dissolution patterns upon addition of fresh water. The hydraulic experiments were conducted using interstitial fluid as well as fresh water. A series of one-step outflow experiments were used to evaluate the drainage patterns for salt-bed heights ranging from 1' to 8'. Measured data include bulk densities and saturated hydraulic conductivities as a function of

Analysis of risks, environmental effects, process feasibility, and costs for disposal of immobilized high-levelwastes in geologic repositories indicates that the disposal system safety has a low sensitivity to the choice of the waste disposal form.

Presents a method to calculate the amount of high-level radioactive waste by taking into consideration the following factors: the fission process that yields the waste, identification of the waste, the energy required to run a 1-GWe plant for one year, and the uranium mass required to produce that energy. Briefly discusses waste disposal and…

The High-LevelWaste System is a set of six different processes interconnected by pipelines. These processes function as one large treatment plant that receives, stores, and treats high-levelwastes from various generators at SRS and converts them into forms suitable for final disposal. The three major forms are borosilicate glass, which will be eventually disposed of in a Federal Repository, Saltstone to be buried on site, and treated water effluent that is released to the environment.

If society is ever to reap the potential benefits of nuclear energy, technologists must close the fuel-cycle completely. A closed cycle equates to a continued supply of fuel and safe reactors, but also reliable and comprehensive closure of waste issues. Highlevelwaste (HLW) disposal in borosilicate glass (BSG) is based on 1970s era evaluations. This host matrix is very adaptable to sequestering a wide variety of radionuclides found in raffinates from spent fuel reprocessing. However, it is now known that the current system is far from optimal for disposal of the diverse HLW streams, and proven alternatives are available to reduce costs by billions of dollars. The basis for HLW disposal should be reassessed to consider extensive waste form and process technology research and development efforts, which have been conducted by the United States Department of Energy (USDOE), international agencies and the private sector. Matching the waste form to the waste chemistry and using currently available technology could increase the waste content in waste forms to 50% or more and double processing rates. Optimization of the HLW disposal system would accelerate HLW disposition and increase repository capacity. This does not necessarily require developing new waste forms, the emphasis should be on qualifying existing matrices to demonstrate protection equal to or better than the baseline glass performance. Also, this proposed effort does not necessarily require developing new technology concepts. The emphasis is on demonstrating existing technology that is clearly better (reliability, productivity, cost) than current technology, and justifying its use in future facilities or retrofitted facilities. Higher waste processing and disposal efficiency can be realized by performing the engineering analyses and trade-studies necessary to select the most efficient methods for processing the full spectrum of wastes across the nuclear complex. This paper will describe technologies being

The purpose of this plan is to document the integrated technology program plan for the Savannah River Site (SRS) High-LevelWaste (HLW) Management System. The mission of the SRS HLW System is to receive and store SRShigh-levelwastes in a see and environmentally sound, and to convert these wastes into forms suitable for final disposal. These final disposal forms are borosilicate glass to be sent to the Federal Repository, Saltstone grout to be disposed of on site, and treated waste water to be released to the environment via a permitted outfall. Thus, the technology development activities described herein are those activities required to enable successful accomplishment of this mission. The technology program is based on specific needs of the SRS HLW System and organized following the systems engineering level 3 functions. Technology needs for each level 3 function are listed as reference, enhancements, and alternatives. Finally, FY-95 funding, deliverables, and schedules are s in Chapter IV with details on the specific tasks that are funded in FY-95 provided in Appendix A. The information in this report represents the vision of activities as defined at the beginning of the fiscal year. Depending on emergent issues, funding changes, and other factors, programs and milestones may be adjusted during the fiscal year. The FY-95 SRS HLW technology program strongly emphasizes startup support for the Defense Waste Processing Facility and In-Tank Precipitation. Closure of technical issues associated with these operations has been given highest priority. Consequently, efforts on longer term enhancements and alternatives are receiving minimal funding. However, High-LevelWaste Management is committed to participation in the national Radioactive Waste Tank Remediation Technology Focus Area. 4 refs., 5 figs., 9 tabs.

The U.S. Department of Energy is putting a modern version of alchemy to work to produce an answer to a decades-old problem. It is taking place at the Savannah River Site (SRS) in Aiken, South Carolina and at the West Valley Demonstration Project (WVDP) near Buffalo, New York. At both locations, contractor Westinghouse Electric Corporation is applying technology that is turning liquid high-level radioactive waste (HLW) into a stabilized, durable glass for safer and easier management. The process is called vitrification. SRS and WVDP are now operating the nation`s first full-scale HLW vitrification plants.

This paper discusses the engineering systems for the structural design of the Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS). At the DWPF, highlevel radioactive liquids will be mixed with glass particles and heated in a melter. This molten glass will then be poured into stainless steel canisters where it will harden. This process will transform the highlevelwaste into a more stable, manageable substance. This paper discuss the structural design requirements for this unique one of a kind facility. A special emphasis will be concentrated on the design criteria pertaining to earthquake, wind and tornado, and flooding.

Researchers successfully demonstrated the chemistry and process equipment of the Caustic-Side Solvent Extraction (CSSX) flowsheet for the decontamination of highlevelwaste using a 33-stage, 2-cm centrifugal contactor apparatus at the Savannah River Technology Center. This represents the first CSSX process demonstration using Savannah River Site (SRS) highlevelwaste. Three tests lasting 6, 12, and 48 hours processed simulated average SRSwaste, simulated Tank 37H/44F composite waste, and Tank 37H/44F highlevelwaste, respectively.

Historically, large items of TRU Waste, which were too large to be packaged in drums for disposal have been packaged in various sizes of custom made plywood boxes at the Savannah River Site (SRS), for many years. These boxes were subsequently packaged into large steel ''Black Boxes'' for storage at SRS, pending availability of Characterization and Certification capability, to facilitate disposal of larger items of TRU Waste. There are approximately 107 Black Boxes in inventory at SRS, each measuring some 18' x 12' x 7', and weighing up to 45,000 lbs. These Black Boxes have been stored since the early 1980s. The project to repackage this waste into Standard Large Boxes (SLBs), Standard Waste Boxes (SWB) and Ten Drum Overpacks (TDOP), for subsequent characterization and WIPP disposal, commenced in FY04. To date, 10 Black Boxes have been repackaged, resulting in 40 SLB-2's, and 37 B25 overpack boxes, these B25's will be overpacked in SLB-2's prior to shipping to WIPP. This paper will describe experience to date from this project.

The national highlevelwaste disposal plans for France, the Federal Republic of Germany, Japan, and the United States are covered. Three conclusions are reached. The first conclusion is that an excellent technology already exists for highlevelwaste disposal. With appropriate packaging, spent fuel seems to be an acceptable waste form. Borosilicate glass reprocessing waste forms are well understood, in production in France, and scheduled for production in the next few years in a number of other countries. For final disposal, a number of candidate geological repository sites have been identified and several demonstration sites opened. The second conclusion is that adequate financing and a legal basis for waste disposal are in place in most countries. Costs of highlevelwaste disposal will probably and about 5 to 10% to the costs of nuclear electric power. Third conclusion is less optimistic.

There are currently about 1055 million curies of high-levelwaste with a thermal output of about 2950 kilowatts (KW) at four sites in the United States: West Valley Demonstration Project (WVDP), Savannah River Site (SRS), Hanford Site (HANF), and Idaho National Engineering Laboratory (INEL). These quantities are expected to increase to about 1200 million curies and 3570 kw by the end of year 2020. Under the Nuclear Waste Policy Act, this high-levelwaste must ultimately be disposed of in a geologic repository. Accordingly, canisters of high-levelwaste immobilized in borosilicate glass or glass-ceramic mixtures are to be produced at the four sites and stored there until a repository becomes available. Data on the estimated production schedules and on the physical, chemical, and radiological characteristics of the canisters of immobilized high-levelwaste have been collected in OCRWM's Waste Characteristics Data Base, including recent updates an revisions. Comparisons of some of these data for the four sites are presented in this report. 14 refs., 3 tabs.

A novel methodology has been developed which makes possible a very fast running computational tool, capable of performing 30 to 50 years of simulation of the entire Savannah River Site (SRS) highlevelwaste complex in less than 2 minutes on a work station. The methodology has been implemented in the Production Planning Model (ProdMod) simulation code which uses Aspen Technology's dynamic simulation software development package SPEEDUP. ProdMod is a pseudo-dynamic simulation code solely based on algebraic equations, using no differential equations. The dynamic nature of the plant process is captured using linear constructs in which the time dependence is implicit. Another innovative approach implemented in ProdMod development is the mapping of event-space on to time-space and vice versa, which accelerates the computation without sacrificing the necessary details in the event-space. ProdMod uses this approach in coupling the time-space continuous simulation with the event-space batch simulation, avoiding the discontinuities inherent in dynamic simulation batch processing. In addition, a general purpose optimization scheme has been devised based on the pseudo-dynamic constructs and the event- and time-space algorithms of ProdMod. The optimization scheme couples a FORTRAN based stand-alone optimization driver with the SPEEDUP based ProdMod simulator to perform dynamic optimization. The scheme is capable of generating single or multiple optimal input conditions for different types of objective functions over single or multiple years of operations depending on the nature of the objective function and operating constraints. The resultant optimal inputs are then interfaced with ProdMod to simulate the dynamic behavior of the waste processing operations. At the conclusion on an optimized advancement step, the simulation parameters are then passed to the optimization driver to generate the next set of optimized parameters. An optimization algorithm using linear programming

At the Savannah River Site (SRS) the Defense Waste Processing Facility (DWPF) has been immobilizing SRS's radioactive highlevelwaste (HLW) sludge into a durable borosilicate glass since 1996. Currently the DWPF has poured over 3,500 canisters, all of which are compliant with the U. S. Department of Energy's (DOE) Waste Acceptance Product Specifications for Vitrified High-LevelWaste Forms (WAPS) and therefore ready to be shipped to a federal geologic repository for permanent disposal. Due to DOE petitioning to withdraw the Yucca Mountain License Application (LA) from the Nuclear Regulatory Commission (NRC) in 2010 and thus no clear disposal path for SRS canistered waste forms, there are opportunities for cost savings with future canister production at DWPF and other DOE producer sites by reevaluating high-levelwaste form requirements and compliance strategies and reducing/eliminating those that will not negatively impact the quality of the canistered waste form.

At the Savannah River Site (SRS) the Defense Waste Processing Facility (DWPF) has been immobilizing SRS's radioactive highlevelwaste (HLW) sludge into a durable borosilicate glass since 1996. Currently the DWPF has poured over 3,500 canisters, all of which are compliant with the U. S. Department of Energy's (DOE) Waste Acceptance Product Specifications for Vitrified High-LevelWaste Forms (WAPS) and therefore ready to be shipped to a federal geologic repository for permanent disposal. Due to DOE petitioning to withdraw the Yucca Mountain License Application (LA) from the Nuclear Regulatory Commission (NRC) in 2010 and thus no clear disposal path for SRS canistered waste forms, there are opportunities for cost savings with future canister production at DWPF and other DOE producer sites by reevaluating high-levelwaste form requirements and compliance strategies and reducing/eliminating those that will not negatively impact the quality of the canistered waste form. (authors)

The primary objective of this study is to demonstrate a mission scenario that uses pairwise and incidental blending of highlevelwaste (HLW) to reduce the total mass of HLW glass. Secondary objectives include understanding how recent refinements to the tank waste inventory and solubility assumptions affect the mass of HLW glass and how logistical constraints may affect the efficacy of HLW blending.

A vitrification facility is being developed by the U.S. Department of Energy (DOE) at the West Valley Demonstration Plant (WVDP) near Buffalo, New York, where approximately 300 canisters of high-level nuclear waste glass will be produced. To assure that the produced waste form is acceptable, uncertainty must be managed. Statistical issues arise due to sampling, waste variations, processing uncertainties, and analytical variations. This paper presents elements of a strategy to characterize and manage the uncertainties associated with demonstrating that an acceptable waste form product is achieved. Specific examples are provided within the context of statistical work performed by Pacific Northwest Laboratory (PNL).

At the Hanford Site in Richland, Washington, the path to site cleanup involves vitrification of the majority of the wastes that currently reside in large underground tanks. A Joule-heated glass melter is the equipment of choice for vitrifying the high-level fraction of these wastes. Even though this technology has general national and international acceptance, opportunities may exist to improve or change the technology to reduce the enormous cost of accomplishing the mission of site cleanup. Consequently, the U.S. Department of Energy requested the staff of the Tanks Focus Area to review immobilization technologies, waste forms, and modifications to requirements for solidification of the high-levelwaste fraction at Hanford to determine what aspects could affect cost reductions with reasonable long-term risk. The results of this study are summarized in this report.

This bibliography contains information on high-level radioactive wastes included in the Department of Energy's Energy Data Base from August 1982 through December 1983. These citations are to research reports, journal articles, books, patents, theses, and conference papers from worldwide sources. Five indexes, each preceded by a brief description, are provided: Corporate Author, Personal Author, Subject, Contract Number, and Report Number. 1452 citations.

Tc contamination is found within the DOE complex at those sites whose mission involved extraction of plutonium from irradiated uranium fuel or isotopic enrichment of uranium. At the Hanford Site, chemical separations and extraction processes generated large amounts of highlevel and transuranic wastes that are currently stored in underground tanks. The waste from these extraction processes is currently stored in underground HighLevelWaste (HLW) tanks. However, the chemistry of the HLW in any given tank is greatly complicated by repeated efforts to reduce volume and recover isotopes. These processes ultimately resulted in mixing of waste streams from different processes. As a result, the chemistry and the fate of Tc in HLW tanks are not well understood. This lack of understanding has been made evident in the failed efforts to leach Tc from sludge and to remove Tc from supernatants prior to immobilization. Although recent interest in Tc chemistry has shifted from pretreatment chemistry to waste residuals, both needs are served by a fundamental understanding of Tc chemistry.

Eleven major Department of Energy (DOE) site contractors were chartered by the Assistant Secretary to use a systems engineering approach to develop and evaluate technically defensible cost savings opportunities across the complex. Known as the complex-wide Environmental Management Integration (EMI), this process evaluated all the major DOE waste streams including highlevelwaste (HLW). Across the DOE complex, this waste stream has the highest life cycle cost and is scheduled to take until at least 2035 before all HLW is processed for disposal. Technical contract experts from the four DOE sites that manage highlevelwaste participated in the integration analysis: Hanford, Savannah River Site (SRS), Idaho National Engineering and Environmental Laboratory (INEEL), and West Valley Demonstration Project (WVDP). In addition, subject matter experts from the Yucca Mountain Project and the Tanks Focus Area participated in the analysis. Also, departmental representatives from the US Department of Energy Headquarters (DOE-HQ) monitored the analysis and results. Workouts were held throughout the year to develop recommendations to achieve a complex-wide integrated program. From this effort, the HLW Environmental Management (EM) Team identified a set of programmatic and technical opportunities that could result in potential cost savings and avoidance in excess of $18 billion and an accelerated completion of the HLW mission by seven years. The cost savings, schedule improvements, and volume reduction are attributed to a multifaceted HLW treatment disposal strategy which involves waste pretreatment, standardized waste matrices, risk-based retrieval, early development and deployment of a shipping system for glass canisters, and reasonable, low cost tank closure.

The purpose of this Analysis/Model Report (AMR) is to document the analyses that were done to develop models for radionuclide release from high-levelwaste (HLW) glass dissolution that can be integrated into performance assessment (PA) calculations conducted to support site recommendation and license application for the Yucca Mountain site. This report was developed in accordance with the ''Technical Work Plan for Waste Form Degradation Process Model Report for SR'' (CRWMS M&O 2000a). It specifically addresses the item, ''Defense HighLevelWaste Glass Degradation'', of the product technical work plan. The AP-3.15Q Attachment 1 screening criteria determines the importance for its intended use of the HLW glass model derived herein to be in the category ''Other Factors for the Postclosure Safety Case-Waste Form Performance'', and thus indicates that this factor does not contribute significantly to the postclosure safety strategy. Because the release of radionuclides from the glass will depend on the prior dissolution of the glass, the dissolution rate of the glass imposes an upper bound on the radionuclide release rate. The approach taken to provide a bound for the radionuclide release is to develop models that can be used to calculate the dissolution rate of waste glass when contacted by water in the disposal site. The release rate of a particular radionuclide can then be calculated by multiplying the glass dissolution rate by the mass fraction of that radionuclide in the glass and by the surface area of glass contacted by water. The scope includes consideration of the three modes by which water may contact waste glass in the disposal system: contact by humid air, dripping water, and immersion. The models for glass dissolution under these contact modes are all based on the rate expression for aqueous dissolution of borosilicate glasses. The mechanism and rate expression for aqueous dissolution are adequately understood; the analyses in this AMR were conducted to

This report describes the rationale and presents instructions for a site ranking system (SRS). SRS ranks hazardous chemical and radioactive waste sites by scoring important and readily available factors that influence risk to human health. Using SRS, sites can be ranked for purposes of detailed site investigations. SRS evaluates the relative risk as a combination of potentially exposed population, chemical toxicity, and potential exposure of release from a waste site; hence, SRS uses the same concepts found in a detailed assessment of health risk. Basing SRS on the concepts of risk assessment tends to reduce the distortion of results found in other ranking schemes. More importantly, a clear logic helps ensure the successful application of the ranking procedure and increases its versatility when modifications are necessary for unique situations. Although one can rank sites using a detailed risk assessment, it is potentially costly because of data and resources required. SRS is an efficient approach to provide an order-of-magnitude ranking, requiring only readily available data (often only descriptive) and hand calculations. Worksheets are included to make the system easier to understand and use. 88 refs., 19 figs., 58 tabs.

A low intensity radiation field is generated by the highlevelwaste that is stored within carbon steel lined tanks at the Savannah River Site (SRS). The highest level of radiation damage to the tank walls from gamma and spontaneous neutron emissions is estimated to be less than 1.0E-6 displacements per atom (DPA) for a 100 year exposure to fresh, ``high heat`` SRSwaste assuming continuous replenishment of the radionuclides. This damage level is below the limit for measurable radiation damage to the mechanical properties of carbon steel. Structural assessment of tanks for storage of highlevelwaste may be based on nominal or code values of the mechanical properties of the steels from which the tanks were constructed.

This report presents evaluations of several methods for the in-process decontamination of metallic canisters containing any one of a number of solidified high-levelwaste (HLW) forms. The use of steam-water, steam, abrasive blasting, electropolishing, liquid honing, vibratory finishing and soaking have been tested or evaluated as potential techniques to decontaminate the outer surfaces of HLW canisters. Either these techniques have been tested or available literature has been examined to assess their applicability to the decontamination of HLW canisters. Electropolishing has been found to be the most thorough method to remove radionuclides and other foreign material that may be deposited on or in the outer surface of a canister during any of the HLW processes. Steam or steam-water spraying techniques may be adequate for some applications but fail to remove all contaminated forms that could be present in some of the HLW processes. Liquid honing and abrasive blasting remove contamination and foreign material very quickly and effectively from small areas and components although these blasting techniques tend to disperse the material removed from the cleaned surfaces. Vibratory finishing is very capable of removing the bulk of contamination and foreign matter from a variety of materials. However, special vibratory finishing equipment would have to be designed and adapted for a remote process. Soaking techniques take long periods of time and may not remove all of the smearable contamination. If soaking involves pickling baths that use corrosive agents, these agents may cause erosion of grain boundaries that results in rough surfaces.

The Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS) in Aiken, SC mg began immobilizing high-level radioactive waste in borosilicate glass in 1996. Currently, the radioactive glass is being produced as a ``sludge-only`` composition by combining washed high-levelwaste sludge with glass frit. The glass is poured in stainless steel canisters which will eventually be disposed of in a permanent, geological repository. To date, DWPF has produced about 100 canisters of vitrified waste. Future processing operations will, be based on a ``coupled`` feed of washed high-levelwaste sludge, precipitated cesium, and glass frit. This paper provides an update of the processing activities completed to date, operational/flowsheet problems encountered, and programs underway to increase production rates.

The information supplied by this 30-year solid waste forecast has been compiled as a source document to the Waste Management Environmental Impact Statement (WMEIS). The WMEIS will help to select a sitewide strategic approach to managing present and future Savannah River Site (SRS) waste generated from ongoing operations, environmental restoration (ER) activities, transition from nuclear production to other missions, and decontamination and decommissioning (D&D) programs. The EIS will support project-level decisions on the operation of specific treatment, storage, and disposal facilities within the near term (10 years or less). In addition, the EIS will provide a baseline for analysis of future waste management activities and a basis for the evaluation of the specific waste management alternatives. This 30-year solid waste forecast will be used as the initial basis for the EIS decision-making process. The Site generates and manages many types and categories of waste. With a few exceptions, waste types are divided into two broad groups-high-levelwaste and solid waste. High-levelwaste consists primarily of liquid radioactive waste, which is addressed in a separate forecast and is not discussed further in this document. The waste types discussed in this solid waste forecast are sanitary waste, hazardous waste, low-level mixed waste, low-level radioactive waste, and transuranic waste. As activities at SRS change from primarily production to primarily decontamination and decommissioning and environmental restoration, the volume of each waste s being managed will change significantly. This report acknowledges the changes in Site Missions when developing the 30-year solid waste forecast.

At the Department of Energy's Savannah River Site, American Recovery and Reinvestment Act funding has supported installation of bubbler technology and related enhancements in the Defense Waste Processing Facility (DWPF). The improvements will accelerate the processing of radioactive waste into a safe, stable form for storage and permit expedited closure of underground waste tanks holding 37 million gallons of liquid nuclear waste.

The objective of the proposed effort is to use a new approach to develop solution models of complex waste glass systems and spent fuel that are predictive with regard to composition, phase separation, and volatility. The effort will also yield thermodynamic values for waste components that are fundamentally required for corrosion models used to predict the leaching/corrosion behavior for waste glass and spent fuel material. This basic information and understanding of chemical behavior can subsequently be used directly in computational models of leaching and transport in geologic media, in designing and engineering waste forms and barrier systems, and in prediction of chemical interactions.

The assessment of release of radionuclides from waste repositories substantially depends on the leaching behavior of the spent fuel or waste form. Assumed rates based on dissolution of specific phases (assumption of unit activity) will lead to potentially grossly overestimated va...

This report presents technical data and performance characteristics of a high-levelwaste glass and canister intended for use in the design of a complete waste encapsulation package suitable for disposal in a geologic repository. The borosilicate glass contained in the stainless steel canister represents the probable type of high-levelwaste product that will be produced in a commercial nuclear-fuel reprocessing plant. Development history is summarized for high-level liquid waste compositions, waste glass composition and characteristics, and canister design. The decay histories of the fission products and actinides (plus daughters) calculated by the ORIGEN-II code are presented.

This document establishes the combination of design and operational configurations that will be used to provide heat removal from high-levelwaste tanks during Phase 1 waste feed delivery to prevent the waste temperature from exceeding tank safety requirement limits. The chosen method--to use the primary and annulus ventilation systems to remove heat from the high-levelwaste tanks--is documented herein.

A new process, using hybrid microwave energy, has been developed as part of the Strategic Research and Development program and successfully applied to treatment of a wide variety of non-radioactive materials, representative of SRS transuranic (TRU) and mixed wastes. Over 35 simulated (non-radioactive) TRU and mixed waste materials were processed individually, as well as in mixed batches, using hybrid microwave energy, a new technology now being patented by Westinghouse Savannah River Company (WSRC).

Optimal Waste Loading (OWL) models have been developed as multipurpose tools for high-levelwaste studies for the Tank Waste Remediation Program at Hanford. Using nonlinear programming techniques, these models maximize the waste loading of the vitrified waste and optimize the glass formers composition such that the glass produced has the appropriate properties within the melter, and the resultant vitrified waste form meets the requirements for disposal. The OWL model can be used for a single waste stream or for blended streams. The models can determine optimal continuous blends or optimal discrete blends of a number of different wastes. The OWL models have been used to identify the most restrictive constraints, to evaluate prospective waste pretreatment methods, to formulate and evaluate blending strategies, and to determine the impacts of variability in the wastes. The OWL models will be used to aid in the design of frits and the maximize the waste in the glass for High-LevelWaste (HLW) vitrification.

The addition of a small amount of reducing agent to a mixture of a high-level radioactive waste calcine and glass frit before the mixture is melted will produce a more homogeneous glass which is leach-resistant and suitable for long-term storage of high-level radioactive waste products.

Discusses the radioactive waste management in Belgium, Canada, France, Germany, India, Italy, Japan, the United Kingdom, the United States, and the USSR. Indicates that scientists and statesmen should look beyond their own lifetimes into future centuries and millennia to conduct long-range plans essential to protection of future generations. (CC)

This revision of the High-LevelWaste (HLW) System Plan aligns SRS HLW program planning with the DOE Savannah River (DOE-SR) Ten Year Plan (QC-96-0005, Draft 8/6), which was issued in July 1996. The objective of the Ten Year Plan is to complete cleanup at most nuclear sites within the next ten years. The two key principles of the Ten Year Plan are to accelerate the reduction of the most urgent risks to human health and the environment and to reduce mortgage costs. Accordingly, this System Plan describes the HLW program that will remove HLW from all 24 old-style tanks, and close 20 of those tanks, by 2006 with vitrification of all HLW by 2018. To achieve these goals, the DWPF canister production rate is projected to climb to 300 canisters per year starting in FY06, and remain at that rate through the end of the program in FY18, (Compare that to past System Plans, in which DWPF production peaked at 200 canisters per year, and the program did not complete until 2026.) An additional $247M (FY98 dollars) must be made available as requested over the ten year planning period, including a one-time $10M to enhance Late Wash attainment. If appropriate resources are made available, facility attainment issues are resolved and regulatory support is sufficient, then completion of the HLW program in 2018 would achieve a $3.3 billion cost savings to DOE, versus the cost of completing the program in 2026. Facility status information is current as of October 31, 1996.

The objective of this document is to provide information on available issued documents that will assist interested parties in finding available data on high-levelwaste and transuranic waste feed compositions, properties, behavior in candidate processing operations, and behavior on candidate product glasses made from those wastes. This initial compilation is only a partial list of available references.

The Savannah River Site Tank Farms have 45 active underground waste tanks used to store and process nuclear waste materials. There are 4 different tank types, ranging in capacity from 2839 m3 to 4921 m3 (750,000 to 1,300,000 gallons). Eighteen of the tanks are older style and do not meet all current federal standards for secondary containment. The older style tanks are the initial focus of waste removal efforts for tank closure and are referred to as closure tanks. Of the original 51 underground waste tanks, six of the original 24 older style tanks have completed waste removal and are filled with grout. The insoluble waste fraction that resides within most waste tanks at SRS requires vigorous agitation to suspend the solids within the waste liquid in order to transfer this material for eventual processing into glass filled canisters at the Defense Waste Processing Facility (DWPF). SRS suspends the solid waste by use of recirculating mixing pumps. Older style tanks generally have limited riser openings which will not support larger mixing pumps, since the riser access is typically 58.4 cm (23 inches) in diameter. Agitation for these tanks has been provided by four long shafted standard slurry pumps (SLP) powered by an above tank 112KW (150 HP) electric motor. The pump shaft is lubricated and cooled in a pressurized water column that is sealed from the surrounding waste in the tank. Closure of four waste tanks has been accomplished utilizing long shafted pump technology combined with heel removal using multiple technologies. Newer style waste tanks at SRS have larger riser openings, allowing the processing of waste solids to be accomplished with four large diameter SLPs equipped with 224KW (300 HP) motors. These tanks are used to process the waste from closure tanks for DWPF. In addition to the SLPs, a 224KW (300 HP) submersible mixer pump (SMP) has also been developed and deployed within older style tanks. The SMPs are product cooled and

The High-Level Radioactive Waste Transportation Handbook serves as a reference to which state officials and members of the general public may turn for information on radioactive waste transportation and on the federal government`s system for transporting this waste under the Civilian Radioactive Waste Management Program. The Handbook condenses and updates information contained in the Midwestern High-Level Radioactive Waste Transportation Primer. It is intended primarily to assist legislators who, in the future, may be called upon to enact legislation pertaining to the transportation of radioactive waste through their jurisdictions. The Handbook is divided into two sections. The first section places the federal government`s program for transporting radioactive waste in context. It provides background information on nuclear waste production in the United States and traces the emergence of federal policy for disposing of radioactive waste. The second section covers the history of radioactive waste transportation; summarizes major pieces of legislation pertaining to the transportation of radioactive waste; and provides an overview of the radioactive waste transportation program developed by the US Department of Energy (DOE). To supplement this information, a summary of pertinent federal and state legislation and a glossary of terms are included as appendices, as is a list of publications produced by the Midwestern Office of The Council of State Governments (CSG-MW) as part of the Midwestern High-Level Radioactive Waste Transportation Project.

Seven candidate waste forms being developed under the direction of the Department of Energy's National High-LevelWaste (HLW) Technology Program, were evaluated as potential media for the immobilization and geologic disposal of high-level nuclear wastes. The evaluation combined preliminary waste form evaluations conducted at DOE defense waste-sites and independent laboratories, peer review assessments, a product performance evaluation, and a processability analysis. Based on the combined results of these four inputs, two of the seven forms, borosilicate glass and a titanate based ceramic, SYNROC, were selected as the reference and alternative forms for continued development and evaluation in the National HLW Program. Both the glass and ceramic forms are viable candidates for use at each of the DOE defense waste-sites; they are also potential candidates for immobilization of commercial reprocessing wastes. This report describes the waste form screening process, and discusses each of the four major inputs considered in the selection of the two forms.

This papers deals with the optimization of the experimental conditions for the estimation of {sup 237}Np in spent-fuel dissolver/high-levelwaste solutions using thenoyltrifluoroacetone as the extractant. (authors)

Cermets are being developed as an alternate method for the fixation of defense and commercial highlevel radioactive waste in a terminal disposal form. Following initial feasibility assessments of this waste form, consisting of ceramic particles dispersed in an iron-nickel base alloy, significantly improved processing methods were developed. The characterization of cermets has continued through property determinations on samples prepared by various methods from a variety of simulated and actual high-levelwastes. This report describes the status of development of the cermet waste form as it has evolved since 1977. 6 tables, 18 figures.

Transuranic (TRU) solid waste that has been generated as a result of the production of nuclear material for the United States defense program at the Savannah River Site (SRS) has been stored in more than 30,000 55-gallon drums and carbon steel boxes since 1953. Nearly two thirds of those containers have been processed and shipped to the Waste Isolation Pilot Plant. Among the containers assayed so far, the results indicate several drums with fissile inventories significantly higher (600-1000 fissile grams equivalent (FGE) {sup 239}Pu) than their original assigned values. While part of this discrepancy can be attributed to the past limited assay capabilities, human errors are believed to be the primary contributor. This paper summarizes the application of nonspherical fissile material configuration in waste containers, resulting in less restrictive mass and spacing limits, increased storage capacity, and several administrative controls for handling and storage of waste containers being modified without compromising safety.

This paper presents results of the radioactive demonstration of the Precipitate Hydrolysis Process (PHP) that will be performed in the Defense Waste Processing Facility (DWPF) at the Savannah River Site. The PHP destroys the tetraphenylborate precipitate that is used at SRS to isolate Cs-137 from caustic High-LevelWaste (HLW) supernates. This process is necessary to decrease the amount of organic compounds going to the melter in the DWPF. Actual radioactive precipitate containing Cs-137 was used for this demonstration.

A solution of launching high-level nuclear waste into space is suggested. Disposal in space includes solidifying the wastes, embedding them in an explosion-proof vehicle, and launching it into earth orbit, and then into a solar orbit. The benefits of such a system include not only the safe disposal of high-levelwaste but also the establishment of an infrastructure for large-scale space exploration and development. Particular attention is given to the wide range of technical choices along with the societal, economic, and political factors needed for success.

Waste streams planned for generation by the Global Nuclear Energy Partnership (GNEP) and existing radioactive HighLevelWaste (HLW) streams containing organic compounds such as the Tank 48H waste stream at Savannah River Site have completed simulant and radioactive testing, respectfully, by Savannah River National Laboratory (SRNL). GNEP waste streams will include up to 53 wt% organic compounds and nitrates up to 56 wt%. Decomposition of high nitrate streams requires reducing conditions, e.g. provided by organic additives such as sugar or coal, to reduce NOX in the off-gas to N2 to meet Clean Air Act (CAA) standards during processing. Thus, organics will be present during the waste form stabilization process regardless of the GNEP processes utilized and exists in some of the highlevel radioactive waste tanks at Savannah River Site and Hanford Tank Farms, e.g. organics in the feed or organics used for nitrate destruction. Waste streams containing high organic concentrations cannot be stabilized with the existing HLW Best Developed Available Technology (BDAT) which is HLW vitrification (HLVIT) unless the organics are removed by pretreatment. The alternative waste stabilization pretreatment process of Fluidized Bed Steam Reforming (FBSR) operates at moderate temperatures (650-750 C) compared to vitrification (1150-1300 C). The FBSR process has been demonstrated on GNEP simulated waste and radioactive waste containing high organics from Tank 48H to convert organics to CAA compliant gases, create no secondary liquid waste streams and create a stable mineral waste form.

Highlevel radioactive wastes are being vitrified at the Savannah River Site for long term disposal. Many of the wastes contain sulfate at concentrations that can be difficult to retain in borosilicate glass. This study involves efforts to optimize the composition of a glass frit for combination with the waste to improve sulfate retention while meeting other process and product performance constraints. The fabrication and characterization of several series of simulated waste glasses are described. The experiments are detailed chronologically, to provide insight into part of the engineering studies used in developing frit compositions for an operating highlevelwaste vitrification facility. The results lead to the recommendation of a specific frit composition and a concentration limit for sulfate in the glass for the next batch of sludge to be processed at Savannah River.

The purpose of ''Design Analysis for the Defense High-LevelWaste Disposal Container'' analysis is to technically define the defense high-levelwaste (DHLW) disposal container/waste package using the Waste Package Department's (WPD) design methods, as documented in ''Waste Package Design Methodology Report'' (CRWMS M&O [Civilian Radioactive Waste Management System Management and Operating Contractor] 2000a). The DHLW disposal container is intended for disposal of commercial high-levelwaste (HLW) and DHLW (including immobilized plutonium waste forms), placed within disposable canisters. The U.S. Department of Energy (DOE)-managed spent nuclear fuel (SNF) in disposable canisters may also be placed in a DHLW disposal container along with HLW forms. The objective of this analysis is to demonstrate that the DHLW disposal container/waste package satisfies the project requirements, as embodied in Defense HighLevelWaste Disposal Container System Description Document (SDD) (CRWMS M&O 1999a), and additional criteria, as identified in Waste Package Design Sensitivity Report (CRWMS M&Q 2000b, Table 4). The analysis briefly describes the analytical methods appropriate for the design of the DHLW disposal contained waste package, and summarizes the results of the calculations that illustrate the analytical methods. However, the analysis is limited to the calculations selected for the DHLW disposal container in support of the Site Recommendation (SR) (CRWMS M&O 2000b, Section 7). The scope of this analysis is restricted to the design of the codisposal waste package of the Savannah River Site (SRS) DHLW glass canisters and the Training, Research, Isotopes General Atomics (TRIGA) SNF loaded in a short 18-in.-outer diameter (OD) DOE standardized SNF canister. This waste package is representative of the waste packages that consist of the DHLW disposal container, the DHLW/HLW glass canisters, and the DOE-managed SNF in disposable canisters. The intended use of this analysis is to

The Department of Energy, in accord with recommendations from the Du Pont Company, has started construction of a Defense Waste Processing Facility (DWPF) at the Savannah River Plant. The facility should be completed by the end of 1988, and full-scale operation should begin in 1990. This facility will immobilize in borosilicate glass the large quantity of high-level radioactive waste now stored at the plant plus the waste to be generated from continued chemical reprocessing operations. The existing wastes at the Savannah River Plant will be completely converted by about 2010. 21 figures.

Sodium chromate has been an effective corrosion inhibitor for the cooling coil systems in Savannah River Site (SRS) waste tanks for over 40 years. Due to age and operating history, cooling coils occasionally fail allowing chromated water to leak into the environment. The costs of reporting and cleaning up chromate spills became significant enough for SRS to consider alternate inhibitors. Confirmatory tests were performed to assess the effectiveness of three alternative corrosion inhibitor systems for the waste tank cooling water systems: (1) sodium molybdate (250 ppm as Mo)/sodium hydroxide (pH 10), (2) sodium molybdate (50 ppm as Mo)/sodium silicate (50 ppm as Si), and (3) sodium nitrite (500 ppm)/ sodium hydroxide (0.01 M). The tests were conducted under stagnant conditions to simulate a worst-case scenario. The results showed that these inhibitors were as effective as chromate at minimizing general corrosion at solution temperatures between 30--70 C. However, the initiation of localized attack in crevice regions, in solutions containing the alternative inhibitors at 70 C, was observed. Also, for the nitrite and the molybdate systems to be effective, suitable biocide is needed. On the other hand, interval coupon tests showed that the molybdate inhibitor systems prevented significant propagation of the localized attack.

The primary purpose of this document is to describe the overall process control strategy for monitoring and controlling the functions associated with the Phase 1B high-levelwaste feed delivery. This document provides the basis for process monitoring and control functions and requirements needed throughput the double-shell tank system during Phase 1 high-levelwaste feed delivery. This document is intended to be used by (1) the developers of the future Process Control Plan and (2) the developers of the monitoring and control system.

Storage of power reactor spent fuel is necessary at present because of the lack of reprocessing operations particularly in the U.S. By considering the above solidification and storage scenario, there is more than reasonable assurance that acceptable, stable, low heat generation rate, solidified waste can be produced, and safely disposed. The public perception of no waste disposal solutions is being exploited by detractors of nuclear power application. The inability to even point to one overall system demonstration lends credibility to the negative assertions. By delaying the gathering of on-line information to qualify repository sites, and to implement a demonstration, the actions of the nuclear power detractors are self serving in that they can continue to point out there is no demonstration of satisfactory high-levelwaste disposal. By maintaining the liquid and solidified high-levelwaste in secure above ground storage until acceptable decay heat generation rates are achieved, by producing a compatible, high integrity, solid waste form, by providing a second or even third barrier as a compound container and by inserting the enclosed waste form in a qualified repository with spacing to assure moderately low temperature disposal conditions, there appears to be no technical reason for not progressing further with the disposal of high-levelwastes and needed implementation of the complete nuclear power fuel cycle.

Radioactive waste is stored in underground storage tanks at the Department of Energy (DOE) Savannah River Site (SRS). The waste tanks store supernatant liquid salts, consisting primarily of sodium nitrate, sodium nitrite, sodium hydroxide, and sludge. An assessment of the potential degradation mechanisms of the highlevelwaste (HLW) tanks determined that nitrate- induced pitting corrosion and stress corrosion cracking were the two most significant degradation mechanisms. Controls on the solution chemistry (minimum nitrite and hydroxide concentrations) are in place to prevent the initiation and propagation of pitting and stress corrosion cracking in the tanks. These controls are based upon a series of experiments performed using simulated solutions on materials used for construction of the tanks. The technical bases and evolution of these controls is presented.

The Defense HighLevelWaste Disposal Container System supports the confinement and isolation of waste within the Engineered Barrier System of the Monitored Geologic Repository (MGR). Disposal containers are loaded and sealed in the surface waste handling facilities, transferred to the underground through the accesses using a rail mounted transporter, and emplaced in emplacement drifts. The defense highlevelwaste (HLW) disposal container provides long-term confinement of the commercial HLW and defense HLW (including immobilized plutonium waste forms [IPWF]) placed within disposable canisters, and withstands the loading, transfer, emplacement, and retrieval loads and environments. US Department of Energy (DOE)-owned spent nuclear fuel (SNF) in disposable canisters may also be placed in a defense HLW disposal container along with commercial HLW waste forms, which is known as co-disposal. The Defense HighLevelWaste Disposal Container System provides containment of waste for a designated period of time, and limits radionuclide release. The disposal container/waste package maintains the waste in a designated configuration, withstands maximum handling and rockfall loads, limits the individual canister temperatures after emplacement, resists corrosion in the expected handling and repository environments, and provides containment of waste in the event of an accident. Defense HLW disposal containers for HLW disposal will hold up to five HLW canisters. Defense HLW disposal containers for co-disposal will hold up to five HLW canisters arranged in a ring and one DOE SNF canister inserted in the center and/or one or more DOE SNF canisters displacing a HLW canister in the ring. Defense HLW disposal containers also will hold two Multi-Canister Overpacks (MCOs) and two HLW canisters in one disposal container. The disposal container will include outer and inner cylinders, outer and inner cylinder lids, and may include a canister guide. An exterior label will provide a means by

Over the past 10 years, the Hanford Site has been transitioning from nuclear materials production to Site cleanup operations. High-levelwaste characterization at the Hanford Site provides data to support present waste processing operations, tank safety programs, and future waste disposal programs. Quality elements in the high-levelwaste characterization program will be presented by following a sample through the data quality objective, sampling, laboratory analysis and data review process. Transition from production to cleanup has resulted in changes in quality systems and program; the changes, as well as other issues in these quality programs, will be described. Laboratory assessment through quality control and performance evaluation programs will be described, and data assessments in the laboratory and final reporting in the tank characterization reports will be discussed.

Plans for the nation's first high-level nuclear waste repository have called for permanently closing and sealing the repository soon after it is filled. However, the hydrologic environment of the proposed site at Yucca Mountain, Nevada, should allow the repository to be kept open and the waste retrievable indefinitely. This would allow direct monitoring of the repository and maintain the options for future generations to improve upon the disposal methods or use the uranium in the spent fuel as an energy resource.

The Savannah River Site (SRS) has 49 underground storage tanks used to store HighLevelWaste (HLW). The tank space in these tanks must be managed to support the continued operation of key facilities. The reduction of the tank volumes in these tanks are accomplished through the use of three atmospheric pressure HLW evaporators. For a decade, evaporation of highly alkaline HLW containing aluminum and silicates has produced sodium aluminosilicate scales causing both operation and criticality hazards in the 2H Evaporator System. Segregation of aluminum-rich wastes from silicate-rich wastes minimizes the amount of scale produced and reduces cleaning expenses, but does not eliminate the scaling nor increases operation flexibility in waste process. Similar issues have affected the aluminum refining industry for many decades. Over the past several years, successful commercial products have been identified to eliminate aluminosilicate fouling in the aluminum industry, but have not been utilized in a nuclear environment. Laboratory quantities of three proprietary aluminosilicate scale inhibitors have been produced and been shown to prevent formation of scales. SRNL has been actively testing these potential inhibitors to examine their radiation stability, radiolytic degradation behaviors, and downstream impacts to determine their viability within the HLW system. One of the tested polymers successfully meets the established criteria for application in the nuclear environment. This paper will describe a summary of the methodology used to prioritize laboratory testing protocols based on potential impacts/risks identified for inhibitor deployment at SRS.

The Savannah River Site (SRS) has 49 underground storage tanks used to store HighLevelWaste (HLW). The tank space in these tanks must be managed to support the continued operation of key facilities. The reduction of the tank volumes in these tanks are accomplished through the use of three atmospheric pressure HLW evaporators. For a decade, evaporation of highly alkaline HLW containing dissolved aluminate and silicate has produced sodium aluminosilicate scales causing both operation and criticality hazards in the 2H Evaporator System. Segregation of aluminum-rich wastes from silicate-rich wastes minimizes the amount of scale produced and reduces cleaning expenses, but does not eliminate the scaling nor increases operation flexibility in waste process. Similar issues have affected the aluminum refining industry for many decades. Over the past several years, successful commercial products have been identified to eliminate aluminosilicate fouling in the aluminum industry, but have not been utilized in a nuclear environment. Laboratory quantities of three proprietary aluminosilicate scale inhibitors have been produced and been shown to prevent formation of scales. SRNL has been actively testing these potential inhibitors to examine their radiation stability, radiolytic degradation behaviors, and downstream impacts to determine their viability within the HLW system. One of the tested polymers successfully meets the established criteria for application in the nuclear environment. This paper will describe a summary of the methodology used to prioritize laboratory testing protocols based on potential impacts/risks identified for inhibitor deployment at SRS. (authors)

With the increasing demand for the development of more nuclear power comes the responsibility to address the technical challenges of immobilizing high-level nuclear wastes in stable solid forms for interim storage or disposition in geologic repositories. The immobilization of high-level nuclear wastes has been an active area of research and development for over 50 years. Borosilicate glasses and complex ceramic composites have been developed to meet many technical challenges and current needs, although regulatory issues, which vary widely from country to country, have yet to be resolved. Cooperative international programs to develop advanced proliferation-resistant nuclear technologies to close the nuclear fuel cycle and increase the efficiency of nuclear energy production might create new separation waste streams that could demand new concepts and materials for nuclear waste immobilization. This article reviews the current state-of-the-art understanding regarding the materials science of glasses and ceramics for the immobilization of high-level nuclear waste and excess nuclear materials and discusses approaches to address new waste streams.

A transient heat balance model was developed to assess the impact of a Submersible Mixer Pump (SMP) on waste temperature during the process of waste mixing and removal for the Type-I Savannah River Site (SRS) tanks. The model results will be mainly used to determine the SMP design impacts on the waste tank temperature during operations and to develop a specification for a new SMP design to replace existing long-shaft mixer pumps used during waste removal. The model will also be used to provide input to the operation planning. This planning will be used as input to pump run duration in order to maintain temperature requirements within the tank during SMP operation. The analysis model took a parametric approach. A series of the modeling analyses was performed to examine how submersible mixer pumps affect tank temperature during waste removal operation in the Type-I tank. The model domain included radioactive decay heat load, two SMP's, and one Submersible Transfer Pump (STP) as heat source terms. The present model was benchmarked against the test data obtained by the tank measurement to examine the quantitative thermal response of the tank and to establish the reference conditions of the operating variables under no SMP operation. The results showed that the model predictions agreed with the test data of the waste temperatures within about 10%. Transient modeling calculations for two potential scenarios of sludge mixing and removal operations have been made to estimate transient waste temperatures within a Type-I waste tank. When two 200-HP submersible mixers and 12 active cooling coils are continuously operated in 100-in tank level and 40 C initial temperature for 40 days since the initiation of mixing operation, waste temperature rises about 9 C in 48 hours at a maximum. Sensitivity studies for the key operating variables were performed. The sensitivity results showed that the chromate cooling coil system provided the primary cooling mechanism to remove process

This report examines borosilicate glass as a means of immobilizing high-level radioactive wastes. Borosilicate glass will encapsulate most of the defense and some of the commercial HLW in the US. The resulting waste forms must meet the requirements of the WA-SRD and the WAPS, which include a short term PCT durability test. The waste form producer must report the composition(s) of the borosilicate waste glass(es) produced but can choose the composition(s) to meet site-specific requirements. Although the waste form composition is the primary determinant of durability, the redox state of the glass; the existence, content, and composition of crystals; and the presence of glass-in-glass phase separation can affect durability. The waste glass should be formulated to avoid phase separation regions. The ultimate result of this effort will be a waste form which is much more stable and potentially less mobile than the liquid highlevel radioactive waste is currently.

Nuclear waste at Savannah River Site (SRS) waste tanks consists of three different types of waste forms. They are the lighter salt solutions referred to as supernate, the precipitated salts as salt cake, and heavier fine solids as sludge. The sludge is settled on the tank floor. About half of the residual waste radioactivity is contained in the sludge, which is only about 8 percentage of the total waste volume. Mixing study to be evaluated here for the Salt Disposition Integration (SDI) project focuses on supernate preparations in waste tanks prior to transfer to the Salt Waste Processing Facility (SWPF) feed tank. The methods to mix and blend the contents of the SRS blend tanks were evalutaed to ensure that the contents are properly blended before they are transferred from the blend tank such as Tank 50H to the SWPF feed tank. The work consists of two principal objectives to investigate two different pumps. One objective is to identify a suitable pumping arrangement that will adequately blend/mix two miscible liquids to obtain a uniform composition in the tank with a minimum level of sludge solid particulate in suspension. The other is to estimate the elevation in the tank at which the transfer pump inlet should be located where the solid concentration of the entrained fluid remains below the acceptance criterion (0.09 wt% or 1200 mg/liter) during transfer operation to the SWPF. Tank 50H is a Waste Tank that will be used to prepare batches of salt feed for SWPF. The salt feed must be a homogeneous solution satisfying the acceptance criterion of the solids entrainment during transfer operation. The work described here consists of two modeling areas. They are the mixing modeling analysis during miscible liquid blending operation, and the flow pattern analysis during transfer operation of the blended liquid. The modeling results will provide quantitative design and operation information during the mixing/blending process and the transfer operation of the blended

Flammable gases can be generated in DOE high-levelwaste tanks, including radiolytic hydrogen, and during cesium precipitation from salt solutions, benzene. Under normal operating conditions the potential for deflagration or detonation from these gases is precluded by purging and...

The objective of this task is to identify potential waste streams that could be treated with the Small Column Ion Exchange (SCIX) and perform an initial assessment of the impact of doing so on the High-LevelWaste (HLW) system. Design of the SCIX system has been performed as a backup technology for decontamination of High-LevelWaste (HLW) at the Savannah River Site (SRS). The SCIX consists of three modules which can be placed in risers inside underground HLW storage tanks. The pump and filter module and the ion exchange module are used to filter and decontaminate the aqueous tank wastes for disposition in Saltstone. The ion exchange module contains Crystalline Silicotitanate (CST in its engineered granular form is referred to as IONSIV{reg_sign} IE-911), and is selective for removal of cesium ions. After the IE-911 is loaded with Cs-137, it is removed and the column is refilled with a fresh batch. The grinder module is used to size-reduce the cesium-loaded IE-911 to make it compatible with the sludge vitrification system in the Defense Waste Processing Facility (DWPF). If installed at the SRS, this SCIX would need to operate within the current constraints of the larger HLW storage, retrieval, treatment, and disposal system. Although the equipment has been physically designed to comply with system requirements, there is also a need to identify which waste streams could be treated, how it could be implemented in the tank farms, and when this system could be incorporated into the HLW flowsheet and planning. This document summarizes a preliminary examination of the tentative HLW retrieval plans, facility schedules, decontamination factor targets, and vitrified waste form compatibility, with recommendations for a more detailed study later. The examination was based upon four batches of salt solution from the currently planned disposition pathway to treatment in the SCIX. Because of differences in capabilities between the SRS baseline and SCIX, these four batches were

This report is a review of waste form options for the immobilization of high-level-liquid wastes from the nuclear fuel cycle. This review covers the status of international research and development on waste forms as of May 1979. Although the emphasis in this report is on waste form properties, process parameters are discussed where they may affect final waste form properties. A summary table is provided listing properties of various nuclear waste form options. It is concluded that proposed waste forms have properties falling within a relatively narrow range. In regard to crystalline versus glass waste forms, the conclusion is that either glass of crystalline materials can be shown to have some advantage when a single property is considered; however, at this date no single waste form offers optimum properties over the entire range of characteristics investigated. A long-term effort has been applied to the development of glass and calcine waste forms. Several additional waste forms have enough promise to warrant continued research and development to bring their state of development up to that of glass and calcine. Synthetic minerals, the multibarrier approach with coated particles in a metal matrix, and high pressure-high temperature ceramics offer potential advantages and need further study. Although this report discusses waste form properties, the total waste management system should be considered in the final selection of a waste form option. Canister design, canister materials, overpacks, engineered barriers, and repository characteristics, as well as the waste form, affect the overall performance of a waste management system. These parameters were not considered in this comparison.

This report documents the preliminary needs assessment of high-levelwaste (HLW) tank lay-up requirements and considerations for the Hanford Site, Idaho Naitonal Engeineering and Environmental Lab (INEEL), Savannah River Site (SRS) and Oak Ridge Reservation (ORR). This assessment includes the development of a high-level requirements and considerations list that evolved from work done for the West Valley Demonstration Project (WVDP) earlier in fiscal year (FY) 2001, and is based on individual site conditions and tank retrieval/tank closure schedules. Because schedules are continually subject to change, this assessment is considered preliminary and needs review and validation by the individual sites. The lay-up decision methodology developed for WVDP was based on standard systems engineering principles, and provided a structured framework for producing an effective, technically-defensible lay-up strategy.

A large amount of radioactive waste has been stored safely at the Savannah River and Hanford sites over the past 46 years. The aim of this report is to review the experimental corrosion studies at Savannah River and Hanford with the intention of identifying the types and rates of corrosion encountered and indicate how these data contribute to tank failure predictions. The compositions of the High-LevelWastes, mild steels used in the construction of the waste tanks and degradation-modes particularly stress corrosion cracking and pitting are discussed. Current concerns at the Hanford Site are highlighted.

Within this paper we present a simplified analytical model to provide insight into the key performance measures of a generic disposal system for highlevelwaste within a geological disposal facility. The model assumes a low solubility waste matrix within a corrosion resistant disposal container surrounded by a low permeability buffer. Radionuclides migrate from the disposal area through a porous geosphere to the biosphere and give a radiological dose to a receptor. The system of equations describing the migration is transformed into Laplace space and an approximation used to determine peak values for the radionuclide mass transfer rate entering the biosphere. Results from the model are compared with those from more detailed numerical models for key radionuclides in the UK highlevelwaste inventory. Such an insight model can provide a valuable second line of argument to assist in confirming the results of more detailed models and build confidence in the safety case for a geological disposal facility.

The U.S. Department of Energy (DOE) is currently processing high-levelwaste (HLW) through a Joule-heated melter (JHM) at the Savannah River Site (SRS) and plans to vitrify HLW and Low activity waste (LAW) at the Hanford Site. Over the past few years at the DWPF, work has concentrated on increasing waste throughput. These efforts are continuing with an emphasis on high alumina content feeds. High alumina feeds have presented specific challenges for the JHM technology regarding the ability to increase waste loading yet still maintain product quality and adequate throughput. Alternatively, vitrification technology innovations are also being investigated as a means to increase waste throughput. The Cold Crucible Induction Melter (CCIM) technology affords the opportunity for higher vitrification process temperatures as compared to the current reference JHM technology. Higher process temperatures may allow for higher waste loading and higher melt rate. Glass formulation testing to support melter demonstration testing was recently completed. This testing was specifically aimed at high alumina concentration wastes. Glass composition property models were utilized as a guide for formulation development. Both CCIM and JHM testing will be conducted so glass formulation testing was targeted at both technologies with a goal to significantly increase waste loading without compromising product quality.

Production sources, radiological properties, relative solubilities in waste, and laboratory analysis techniques for the forty-five radionuclides identified in Hanford's Waste Treatment and Immobilization Plant (WTP) Feed Acceptance Data Quality Objectives (DQO) document are addressed in this report. Based on Savannah River Site (SRS) experience and waste characteristics, thirteen of the radionuclides are judged to be candidates for potential scaling in HighLevelWaste (HLW) based on the concentrations of other radionuclides as determined through laboratory measurements. The thirteen radionuclides conducive to potential scaling are: Ni-59, Zr-93, Nb-93m, Cd-113m, Sn-121m, Sn-126, Cs-135, Sm-151, Ra-226, Ra-228, Ac-227, Pa-231, and Th-229. The ability to scale radionuclides is useful from two primary perspectives: 1) it provides a means of checking the radionuclide concentrations that have been determined by laboratory analysis; and 2) it provides a means of estimating radionuclide concentrations in the absence of a laboratory analysis technique or when a complex laboratory analysis technique fails. Along with the rationale for identifying and applying the potential scaling factors, this report also provides examples of using the scaling factors to estimate concentrations of radionuclides in current SRSwaste and into the future. Also included in the report are examples of independent laboratory analysis techniques that can be used to check results of key radionuclide analyses. Effective utilization of radionuclide scaling factors requires understanding of the applicable production sources and the chemistry of the waste. As such, the potential scaling approaches identified in this report should be assessed from the perspective of the Hanford waste before reaching a decision regarding WTP applicability.

Thermal design constraints exist on the processing operations in the HighLevelWaste (HLW) tanks of the Savannah River Site (SRS). A FORTRAN computer code was developed to provide a simple, fast, and reasonably accurate analysis tool for plant operation design. The code computes a lumped transient temperature for the liquid contents of a waste tank by modeling the liquid (slurry), the vapor space above it, the tank wall, and the cooling air outside of the tank. Results for a typical processing cycle of several months` duration can be obtained in 2--4 minutes CPU time on a VAX computer. This paper discusses the code`s mathematical models, presents model results for a typical HLW process schedule, and compares the code predictions with operations data.

The Defense HighLevelWaste Disposal Container System supports the confinement and isolation of waste within the Engineered Barrier System of the Monitored Geologic Repository (MGR). Disposal containers are loaded and sealed in the surface waste handling facilities, transferred to the underground through the accesses using a rail mounted transporter, and emplaced in emplacement drifts. The defense highlevelwaste (HLW) disposal container provides long-term confinement of the commercial HLW and defense HLW (including immobilized plutonium waste forms (IPWF)) placed within disposable canisters, and withstands the loading, transfer, emplacement, and retrieval loads and environments. U.S. Department of Energy (DOE)-owned spent nuclear fuel (SNF) in disposable canisters may also be placed in a defense HLW disposal container along with commercial HLW waste forms, which is known as 'co-disposal'. The Defense HighLevelWaste Disposal Container System provides containment of waste for a designated period of time, and limits radionuclide release. The disposal container/waste package maintains the waste in a designated configuration, withstands maximum handling and rockfall loads, limits the individual canister temperatures after emplacement, resists corrosion in the expected handling and repository environments, and provides containment of waste in the event of an accident. Defense HLW disposal containers for HLW disposal will hold up to five HLW canisters. Defense HLW disposal containers for co-disposal will hold up to five HLW canisters arranged in a ring and one DOE SNF canister in the ring. Defense HLW disposal containers also will hold two Multi-Canister Overpacks (MCOs) and two HLW canisters in one disposal container. The disposal container will include outer and inner cylinders, outer and inner cylinder lids, and may include a canister guide. An exterior label will provide a means by which to identify the disposal container and its contents. Different materials

SRSHighLevelWaste Tank Farm personnel rely on conductivity probes for detection of incipient overflow conditions in waste tanks. Minimal information is available concerning the sensitivity that must be achieved such that that liquid detection is assured. Overly sensitive electronics results in numerous nuisance alarms for these safety-related instruments. In order to determine the minimum sensitivity required of the probe, Tank Farm Engineering personnel need adequate conductivity data to improve the existing designs. Little or no measurements of liquid waste conductivity exist; however, the liquid phase of the waste consists of inorganic electrolytes for which the conductivity may be calculated. Savannah River Remediation (SRR) Tank Farm Facility Engineering requested SRNL to determine the conductivity of the supernate resident in SRSwaste Tank 40 experimentally as well as computationally. In addition, SRNL was requested to develop a correlation, if possible, that would be generally applicable to liquid waste resident in SRSwaste tanks. A waste sample from Tank 40 was analyzed for composition and electrical conductivity as shown in Table 4-6, Table 4-7, and Table 4-9. The conductivity for undiluted Tank 40 sample was 0.087 S/cm. The accuracy of OLI Analyzer™ was determined using available literature data. Overall, 95% of computed estimates of electrical conductivity are within ±15% of literature values for component concentrations from 0 to 15 M and temperatures from 0 to 125 °C. Though the computational results are generally in good agreement with the measured data, a small portion of literature data deviates as much as ±76%. A simplified model was created that can be used readily to estimate electrical conductivity of waste solution in computer spreadsheets. The variability of this simplified approach deviates up to 140% from measured values. Generally, this model can be applied to estimate the conductivity within a factor of two. The comparison of the

This publication is intended to provide its readers with an introduction to the issues surrounding the subject of transportation of spent nuclear fuel and high-level radioactive waste, especially as those issues impact the southern region of the United States. It was originally issued by the Southern States Energy Board (SSEB) in July 1987 as the Spent Nuclear Fuel and High-Level Radioactive Waste Transportation Primer, a document patterned on work performed by the Western Interstate Energy Board and designed as a ``comprehensive overview of the issues.`` This work differs from that earlier effort in that it is designed for the educated layman with little or no background in nuclear waste issues. In addition, this document is not a comprehensive examination of nuclear waste issues but should instead serve as a general introduction to the subject. Owing to changes in the nuclear waste management system, program activities by the US Department of Energy and other federal agencies and developing technologies, much of this information is dated quickly. While this report uses the most recent data available, readers should keep in mind that some of the material is subject to rapid change. SSEB plans periodic updates in the future to account for changes in the program. Replacement pages sew be supplied to all parties in receipt of this publication provided they remain on the SSEB mailing list.

This publication is intended to provide its readers with an introduction to the issues surrounding the subject of transportation of spent nuclear fuel and high-level radioactive waste, especially as those issues impact the southern region of the United States. It was originally issued by the Southern States Energy Board (SSEB) in July 1987 as the Spent Nuclear Fuel and High-Level Radioactive Waste Transportation Primer, a document patterned on work performed by the Western Interstate Energy Board and designed as a ``comprehensive overview of the issues.`` This work differs from that earlier effort in that it is designed for the educated layman with little or no background in nuclear waste issues. In addition, this document is not a comprehensive examination of nuclear waste issues but should instead serve as a general introduction to the subject. Owing to changes in the nuclear waste management system, program activities by the US Department of Energy and other federal agencies and developing technologies, much of this information is dated quickly. While this report uses the most recent data available, readers should keep in mind that some of the material is subject to rapid change. SSEB plans periodic updates in the future to account for changes in the program. Replacement pages will be supplied to all parties in receipt of this publication provided they remain on the SSEB mailing list.

This publication is intended to provide its readers with an introduction to the issues surrounding the subject of transportation of spent nuclear fuel and high-level radioactive waste, especially as those issues impact the southern region of the United States. It was originally issued by SSEB in July 1987 as the Spent Nuclear Fuel and High-Level Radioactive Waste Transportation Primer, a document patterned on work performed by the Western Interstate Energy Board and designed as a ``comprehensive overview of the issues.`` This work differs from that earlier effort in that it is designed for the educated layman with little or no background in nuclear waste Issues. In addition. this document is not a comprehensive examination of nuclear waste issues but should instead serve as a general introduction to the subject. Owing to changes in the nuclear waste management system, program activities by the US Department of Energy and other federal agencies and developing technologies, much of this information is dated quickly. While this report uses the most recent data available, readers should keep in mind that some of the material is subject to rapid change. SSEB plans periodic updates in the future to account for changes in the program. Replacement pages will be supplied to all parties in receipt of this publication provided they remain on the SSEB mailing list.

The TRANSCOM (transportation tracking and communication) system is the U.S. Department of Energy`s (DOE`s) real-time system for tracking shipments of spent fuel, high-levelwastes, and other high-visibility shipments of radioactive material. The TRANSCOM system has been operational since 1988. The system was used during FY1993 to track almost 100 shipments within the US.DOE complex, and it is accessed weekly by 10 to 20 users.

The mixing processes in large, complex enclosures using one-dimensional differential equations, with transport in free and wall jets is modeled using standard integral techniques. With this goal in mind, we have constructed a simple, computationally efficient numerical tool, the Berkeley Mechanistic Mixing Model, which can be used to predict the transient evolution of fuel and oxygen concentrations in DOE high-levelwaste tanks following loss of ventilation, and validate the model against a series of experiments.

Alloys under consideration as candidates for the highlevel nuclear waste containers at Yucca Mountain were exposed to a range of corrosion conditions and their performance measured. The alloys tested were Incoloy 825, 70/30 Copper-Nickel, Monel 400, Hastelloy C- 22, and low carbon steel. The test conditions varied were: temperature, concentration, agitation, and crevice simulation. Only in the case of carbon steel was significant attack noted. This attack appeared to be transport limited.

The Savannah River Site's (SRS) Defense Waste Processing Facility (DWPF) began stabilizing highlevelwaste (HLW) in a glass matrix in 1996. Over the past few years, there have been several process and equipment improvements at the DWPF to increase the rate at which the highlevelwaste can be stabilized. These improvements have either directly increased waste processing rates or have desensitized the process to upsets, thereby minimizing downtime and increasing production. Improvements due to optimization of waste throughput with increased HLW loading of the glass resulted in a 6% waste throughput increase based upon operational efficiencies. Improvements in canister production include the pour spout heated bellows liner (5%), glass surge (siphon) protection software (2%), melter feed pump software logic change to prevent spurious interlocks of the feed pump with subsequent dilution of feed stock (2%) and optimization of the steam atomized scrubber (SAS) operation to minimize downtime (3%) for a total increase in canister production of 12%. A number of process recovery efforts have allowed continued operation. These include the off gas system pluggage and restoration, slurry mix evaporator (SME) tank repair and replacement, remote cleaning of melter top head center nozzle, remote melter internal inspection, SAS pump J-Tube recovery, inadvertent pour scenario resolutions, dome heater transformer bus bar cooling water leak repair and new Infra-red camera for determination of glass height in the canister are discussed.

The SYNROC method for immobilization of high-level nuclear reactor wastes is currently being applied to US defense wastes in tank storage at Savannah River, South Carolina. The minerals zirconolite, perovskite, and hollandite are used in SYNROC D formulations to immobilize fission products and actinides that comprise up to 10% of defense waste sludges and coexisting solutions. Additional phases in SYNROC D are nepheline, the host phase for sodium; and spinel, the host for excess aluminum and iron. Up to 70 wt % of calcined sludge can be incorporated with 30 wt % of SYNROC additives to produce a waste form consisting of 10% nepheline, 30% spinel, and approximately 20% each of the radioactive waste-bearing phases. Urea coprecipitation and spray drying/calcining methods have been used in the laboratory to produce homogeneous, reactive ceramic powders. Hot pressing and sintering at temperatures from 1000 to 1100/sup 0/C result in waste form products with greater than 97% of theoretical density. Hot isostatic pressing has recently been implemented as a processing alternative. Characterization of waste-form mineralogy has been done by means of XRD, SEM, and electron microprobe. Leaching of SYNROC D samples is currently being carried out. Assessment of radiation damage effects and physical properties of SYNROC D will commence in FY 81.

A radioactive highlevelwaste glass was made in 1980 with Savannah River Site (SRS) Tank 15 waste. This glass was buried in the SRS burial ground for 24 years but lysimeter data was only available for the first 8 years. The glass was exhumed and analyzed in 2004. The glass was predicted to be very durable and laboratory tests confirmed the durability response. The laboratory results indicated that the glass was very durable as did analysis of the lysimeter data. Scanning electron microscopy of the glass burial surface showed no significant glass alteration consistent with the results of the laboratory and field tests. No detectable Pu, Am, Cm, Np, or Ru leached from the glass into the surrounding sediment. Leaching of {beta}/{delta} from {sup 90}Sr and {sup 137}Cs in the glass was diffusion controlled. Less than 0.5% of the Cs and Sr in the glass leached into the surrounding sediment, with >99% of the leached radionuclides remaining within 8 centimeters of the glass pellet.

Simulated Savannah River Site (SRS) waste glass forms have been buried in the granite geology of the Stirpa mine in Sweden for two years. Analyses of glass surfaces provided a measure of the performance of the waste glasses as a function of time. Similar SRSwaste glass compositions have also been buried in salt at the WIPP facility in Carlsbad, New Mexico for a similar time period. Analyses of the SRSwaste glasses buried in-situ in granite will be presented and compared to the performance of these same compositions buried in salt at WIPP.

Simulated Savannah River Site (SRS) waste glass forms have been buried in the granite geology of the Stirpa mine in Sweden for two years. Analyses of glass surfaces provided a measure of the performance of the waste glasses as a function of time. Similar SRSwaste glass compositions have also been buried in salt at the WIPP facility in Carlsbad, New Mexico for a similar time period. Analyses of the SRSwaste glasses buried in-situ in granite will be presented and compared to the performance of these same compositions buried in salt at WIPP.

The Defense Waste Processing Facility (DWPF) at the Savannah River Site vitrifies HighLevelWaste (HLW) for repository internment. The process consists of three major steps: waste pretreatment, vitrification, and canister decontamination/sealing. The HLW consists of insoluble metal hydroxides (primarily iron, aluminum, magnesium, manganese, and uranium) and soluble sodium salts (carbonate, hydroxide, nitrite, nitrate, and sulfate). The HLW is processed in large batches through DWPF; DWPF has recently completed processing Sludge Batch 3 (SB3) and is currently processing Sludge Batch 4 (SB4). The composition of metal species in SB4 is shown in Table 1 as a function of the ratio of a metal to iron. Simulants remove radioactive species and renormalize the remaining species. Supernate composition is shown in Table 2.

The Westinghouse Savannah River Technology Center was requested by it`s sister site, West Valley Nuclear Service (WVNS), to develop a remote inspection system to gather wall thickness readings of their HighLevelWaste Tanks. WVNS management chose to take a proactive approach to gain current information on two tanks t hat had been in service since the early 70`s. The tanks contain highlevelwaste, are buried underground, and have only two access ports to an annular space between the tank and the secondary concrete vault. A specialized remote system was proposed to provide both a visual surveillance and ultrasonic thickness measurements of the tank walls. A magnetic wheeled crawler was the basis for the remote delivery system integrated with an off-the-shelf Ultrasonic Data Acquisition System. A development program was initiated for Savannah River Technology Center (SRTC) to design, fabricate, and test a remote system based on the Crawler. The system was completed and involved three crawlers to perform the needed tasks, an Ultrasonic Crawler, a Camera Crawler, and a Surface Prep Crawler. The crawlers were computer controlled so that their operation could be done remotely and their position on the wall could be tracked. The Ultrasonic Crawler controls were interfaced with ABB Amdata`s I-PC, Ultrasonic Data Acquisition System so that thickness mapping of the wall could be obtained. A second system was requested by Westinghouse Savannah River Company (WSRC), to perform just ultrasonic mapping on their similar Waste Storage Tanks; however, the system needed to be interfaced with the P-scan Ultrasonic Data Acquisition System. Both remote inspection systems were completed 9/94. Qualifications tests were conducted by WVNS prior to implementation on the actual tank and tank development was achieved 10/94. The second inspection system was deployed at WSRC 11/94 with success, and the system is now in continuous service inspecting the remaining highlevelwaste tanks at WSRC.

Space disposal of selected components of military high-levelwaste (HLW) is considered. This disposal option offers the promise of eliminating the long-lived radionuclides in military HLW from the earth. A space mission which meets the dual requirements of long-term orbital stability and a maximum of one space shuttle launch per week over a period of 20-40 years, is a heliocentric orbit about halfway between the orbits of earth and Venus. Space disposal of high-level radioactive waste is characterized by long-term predictability and short-term uncertainties which must be reduced to acceptably low levels. For example, failure of either the Orbit Transfer Vehicle after leaving low earth orbit, or the storable propellant stage failure at perihelion would leave the nuclear waste package in an unplanned and potentially unstable orbit. Since potential earth reencounter and subsequent burn-up in the earth's atmosphere is unacceptable, a deep space rendezvous, docking, and retrieval capability must be developed.

The objective of this study was to experimentally measure the properties and performance of a series of glasses with compositions that could represent highlevelwaste Sludge Batch 5 (SB5) as vitrified at the Savannah River Site Defense Waste Processing Facility. These data were used to guide frit optimization efforts as the SB5 composition was finalized. Glass compositions for this study were developed by combining a series of SB5 composition projections with a group of candidate frits. The study glasses were fabricated using depleted uranium and their chemical compositions, crystalline contents and chemical durabilities were characterized. Trevorite was the only crystalline phase that was identified in a few of the study glasses after slow cooling, and is not of concern as spinels have been shown to have little impact on the durability of highlevelwaste glasses. Chemical durability was quantified using the Product Consistency Test (PCT). All of the glasses had very acceptable durability performance. The results of this study indicate that a frit composition can be identified that will provide a processable and durable glass when combined with SB5.

The purpose of this analysis is to document the Quality Assurance (QA) classification of the Monitored Geologic Repository (MGR) defense high-levelwaste disposal container system structures, systems and components (SSCs) performed by the MGR Safety Assurance Department. This analysis also provides the basis for revision of YMP/90-55Q, Q-List (YMP 1998). The Q-List identifies those MGR SSCs subject to the requirements of DOE/RW-0333PY ''Quality Assurance Requirements and Description'' (QARD) (DOE 1998).

Setpoints for nuclear safety-related instrumentation are required for actions determined by the design authorization basis. Minimum requirements need to be established for assuring that setpoints are established and held within specified limits. This document establishes the controlling methodology for changing setpoints of all classifications. The instrumentation under consideration involve the transfer, storage, and volume reduction of radioactive liquid waste in the F- and H-Area High-Level Radioactive Waste Tank Farms. The setpoint document will encompass the PROCESS AREA listed in the Safety Analysis Report (SAR) (DPSTSA-200-10 Sup 18) which includes the diversion box HDB-8 facility. In addition to the PROCESS AREAS listed in the SAR, Building 299-H and the Effluent Transfer Facility (ETF) are also included in the scope.

The socioeconomic investigations of possible impacts of the proposed repository for high-level nuclear waste at Yucca Mountain, Nevada, have been unprecedented in several respects. They bear on the public decision that sooner or later will be made as to where and how to dispose permanently of the waste presently at military weapons installations and that continues to accumulate at nuclear power stations. No final decision has yet been made. There is no clear precedent from other countries. The organization of state and federal studies is unique. The state studies involve more disciplines than any previous efforts. They have been carried out in parallel to federal studies and have pioneered in defining some problems and appropriate research methods. A recent annotated bibliography provides interested scientists with a compact guide to the 178 published reports, as well as to relevant journal articles and related documents. PMID:7971963

The socioeconomic investigations of possible impacts of the proposed repository for high-level nuclear waste at Yucca Mountain, Nevada, have been unprecedented in several respects. They bear on the public decision that sooner or later will be made as to where and how to dispose permanently of the waste presently at military weapons installations and that continues to accumulate at nuclear power stations. No final decision has yet been made. There is no clear precedent from other countries. The organization of state and federal studies is unique. The state studies involve more disciplines than any previous efforts. They have been carried out in parallel to federal studies and have pioneered in defining some problems and appropriate research methods. A recent annotated bibliography provides interested scientists with a compact guide to the 178 published reports, as well as to relevant journal articles and related documents. PMID:7971963

Remote operability and maintainability of vitrification equipment were assessed under shielded-cell conditions. The equipment tested will be applied to immobilize high-level and transuranic liquid waste slurries that resulted from plutonium production for defense weapons. Equipment tested included: a turntable for handling waste canisters under the melter; a removable discharge cone in the melter overflow section; a thermocouple jumper that extends into a shielded cell; remote instrument and electrical connectors; remote, mechanical, and heat transfer aspects of the melter glass overflow section; a reamer to clean out plugged nozzles in the melter top; a closed circuit camera to view the melter interior; and a device to retrieve samples of the glass product. A test was also conducted to evaluate liquid metals for use in a liquid metal sealing system.

A review of the data collected during ultrasonic inspection of the Type I highlevelwaste tanks has been completed. The data was analyzed for relevance to the possibility of vapor space corrosion and liquid/air interface corrosion. The review of the Type I tank UT inspection data has confirmed that the vapor space general corrosion is not an unusually aggressive phenomena and correlates well with predicted corrosion rates for steel exposed to bulk solution. The corrosion rates are seen to decrease with time as expected. The review of the temperature data did not reveal any obvious correlations between high temperatures and the occurrences of leaks. The complex nature of temperature-humidity interaction, particularly with respect to vapor corrosion requires further understanding to infer any correlation. The review of the waste level data also did not reveal any obvious correlations.

The Tank Waste Remediation System (TWRS) Storage and Disposal Project has established the Immobilized High-LevelWaste (IBLW) Storage Sub-Project to provide the capability to store Phase I and II BLW products generated by private vendors. A design/construction project, Project W-464, was established under the Sub-Project to provide the Phase I capability. Project W-464 will retrofit the Hanford Site Canister Storage Building (CSB) to accommodate the Phase I I-ILW products. Project W-464 conceptual design is currently being performed to interim store 3.0 m-long BLW stainless steel canisters with a 0.61 in diameter, DOE is considering using a 4.5 in canister of the same diameter to reduce permanent disposal costs. This study was performed to assess the impact of replacing the 3.0 in canister with the 4.5 in canister. The summary cost and schedule impacts are described.

For more than half a century, the Council of State Governments has served as a common ground for the states of the nation. The Council is a nonprofit, state-supported and -directed service organization that provides research and resources, identifies trends, supplies answers and creates a network for legislative, executive and judicial branch representatives. This List of Available Resources was prepared with the support of the US Department of Energy, Cooperative Agreement No. DE-FC02-89CH10402. However, any opinions, findings, conclusions, or recommendations expressed herein are those of the author(s) and do not necessarily reflect the views of DOE. The purpose of the agreement, and reports issued pursuant to it, is to identify and analyze regional issues pertaining to the transportation of high-level radioactive waste and to inform Midwestern state officials with respect to technical issues and regulatory concerns related to waste transportation.

The Department of Energy has authorized a team of glass formulation and processing experts at the Savannah River National Laboratory (SRNL), the Pacific Northwest National Laboratory (PNNL), and the Vitreous State Laboratory (VSL) at Catholic University of America to develop a systematic approach to increase highlevelwaste melter throughput (by increasing waste loading with minimal or positive impacts on melt rate). This task is aimed at proof-of-principle testing and the development of tools to improve waste loading and melt rate, which will lead to higher waste throughput. Four specific tasks have been proposed to meet these objectives (for details, see WSRC-STI-2007-00483): (1) Integration and Oversight, (2) Crystal Accumulation Modeling (led by PNNL)/Higher Waste Loading Glasses (led by SRNL), (3) Melt Rate Evaluation and Modeling, and (4) Melter Scale Demonstrations. Task 2, Crystal Accumulation Modeling/Higher Waste Loading Glasses is the focus of this report. The objective of this study is to provide supplemental data to support the possible use of alternative melter technologies and/or implementation of alternative process control models or strategies to target higher waste loadings (WLs) for the Defense Waste Processing Facility (DWPF)--ultimately leading to higher waste throughputs and a reduced mission life. The glass selection strategy discussed in this report was developed to gain insight into specific technical issues that could limit or compromise the ability of glass formulation efforts to target higher WLs for future sludge batches at the Savannah River Site (SRS). These technical issues include Al-dissolution, higher TiO{sub 2} limits and homogeneity issues for coupled-operations, Al{sub 2}O{sub 3} solubility, and nepheline formation. To address these technical issues, a test matrix of 28 glass compositions has been developed based on 5 different sludge projections for future processing. The glasses will be fabricated and characterized based on

The U.S. Department of Energy Office of River Protection (ORP) has implemented an integrated program to increase the loading of Hanford tank wastes in glass while meeting melter lifetime expectancies and process, regulatory, and product quality requirements. The integrated ORP program is focused on providing a technical, science-based foundation from which key decisions can be made regarding the successful operation of the Hanford Tank Waste Treatment and Immobilization Plant (WTP) facilities. The fundamental data stemming from this program will support development of advanced glass formulations, key process control models, and tactical processing strategies to ensure safe and successful operations for both the low-activity waste (LAW) and high-levelwaste (HLW) vitrification facilities with an appreciation toward reducing overall mission life. The purpose of this advanced HLW glass research and development plan is to identify the near-, mid-, and longer-term research and development activities required to develop and validate advanced HLW glasses and their associated models to support facility operations at WTP, including both direct feed and full pretreatment flowsheets. This plan also integrates technical support of facility operations and waste qualification activities to show the interdependence of these activities with the advanced waste glass (AWG) program to support the full WTP mission. Figure ES-1 shows these key ORP programmatic activities and their interfaces with both WTP facility operations and qualification needs. The plan is a living document that will be updated to reflect key advancements and mission strategy changes. The research outlined here is motivated by the potential for substantial economic benefits (e.g., significant increases in waste throughput and reductions in glass volumes) that will be realized when advancements in glass formulation continue and models supporting facility operations are implemented. Developing and applying advanced

The current baseline assumption is that packaging ¡§as is¡¨ and direct disposal of highlevelwaste (HLW) calcine in a Monitored Geologic Repository will be allowed. The fall back position is to develop a stabilized waste form for the HLW calcine, that will meet repository waste acceptance criteria currently in place, in case regulatory initiatives are unsuccessful. A decision between direct disposal or a stabilization alternative is anticipated by June 2006. The purposes of this Engineering Design File (EDF) are to provide a pre-conceptual design on three low temperature processes under development for stabilization of highlevelwaste calcine (i.e., the grout, hydroceramic grout, and iron phosphate ceramic processes) and to support a down selection among the three candidates. The key assumptions for the pre-conceptual design assessment are that a) a waste treatment plant would operate over eight years for 200 days a year, b) a design processing rate of 3.67 m3/day or 4670 kg/day of HLW calcine would be needed, and c) the performance of waste form would remove the HLW calcine from the hazardous waste category, and d) the waste form loadings would range from about 21-25 wt% calcine. The conclusions of this EDF study are that: (a) To date, the grout formulation appears to be the best candidate stabilizer among the three being tested for HLW calcine and appears to be the easiest to mix, pour, and cure. (b) Only minor differences would exist between the process steps of the grout and hydroceramic grout stabilization processes. If temperature control of the mixer at about 80„aC is required, it would add a major level of complexity to the iron phosphate stabilization process. (c) It is too early in the development program to determine which stabilizer will produce the minimum amount of stabilized waste form for the entire HLW inventory, but the volume is assumed to be within the range of 12,250 to 14,470 m3. (d) The stacked vessel height of the hot process vessels

Nuclear and commercial non-nuclear technologies that have the potential of meeting the environmental restoration, decontamination and decommissioning, and high-levelwaste management objectives are being assessed and evaluated. A detailed comparison of innovative technologies available will be performed to determine the safest and most economical technology for meeting these objectives. Information derived from this effort will be matched with the multi-objectives of the environmental restoration, decontamination and decommissioning, and high-levelwaste management effort to ensure that the best, most economical, and the safest technologies are used in decision making at USDOE-SRS. Technology-related variables will be developed and the resulting data formatted and computerized for multimedia systems. The multimedia system will be made available to technology developers and evaluators to ensure that the best, most economical, and the safest technologies are used in decision making at USDOE-SRS. Technology-related variables will be developed and the resulting data formatted and computerized for multimedia systems. The multimedia system will be made available to technology developers and evaluators to ensure that the safest and most economical technologies are developed for use at SRS and other DOE sites.

Preliminary evaluation of deep borehole disposal of high-level radioactive waste and spent nuclear fuel indicates the potential for excellent long-term safety performance at costs competitive with mined repositories. Significant fluid flow through basement rock is prevented, in part, by low permeabilities, poorly connected transport pathways, and overburden self-sealing. Deep fluids also resist vertical movement because they are density stratified. Thermal hydrologic calculations estimate the thermal pulse from emplaced waste to be small (less than 20 C at 10 meters from the borehole, for less than a few hundred years), and to result in maximum total vertical fluid movement of {approx}100 m. Reducing conditions will sharply limit solubilities of most dose-critical radionuclides at depth, and high ionic strengths of deep fluids will prevent colloidal transport. For the bounding analysis of this report, waste is envisioned to be emplaced as fuel assemblies stacked inside drill casing that are lowered, and emplaced using off-the-shelf oilfield and geothermal drilling techniques, into the lower 1-2 km portion of a vertical borehole {approx}45 cm in diameter and 3-5 km deep, followed by borehole sealing. Deep borehole disposal of radioactive waste in the United States would require modifications to the Nuclear Waste Policy Act and to applicable regulatory standards for long-term performance set by the US Environmental Protection Agency (40 CFR part 191) and US Nuclear Regulatory Commission (10 CFR part 60). The performance analysis described here is based on the assumption that long-term standards for deep borehole disposal would be identical in the key regards to those prescribed for existing repositories (40 CFR part 197 and 10 CFR part 63).

This report discusses the Accelerator Transmutation of Waste (ATW) concept which aims at destruction of key long-lived radionuclides in high-level nuclear waste (HLW), both fission products and actinides. This focus makes it different from most other transmutation concepts which concentrate primarily on actinide burning. The ATW system uses an accelerator-driven, sub-critical assembly to create an intense thermal neutron environment for radionuclide transmutation. This feature allows rapid transmutation under low-inventory system conditions, which in turn, has a direct impact on the size of chemical separations and materials handling components of the system. Inventories in ATW are factors of eight to thirty times smaller than reactor systems of equivalent thermal power. Chemical separations systems are relatively small in scale and can be optimized to achieve high decontamination factors and minimized waste streams. The low-inventory feature also directly impacts material amounts remaining in the system at its end of life. In addition to its low-inventory operation, the accelerator-driven neutron source features of ATW are key to providing a sufficient level of neutrons to allow transmutation of long-lived fission products.

The Defense High-LevelWaste Leaching Mechanisms Program brought six major US laboratories together for three years of cooperative research. The participants reached a consensus that solubility of the leached glass species, particularly solubility in the altered surface layer, is the dominant factor controlling the leaching behavior of defense waste glass in a system in which the flow of leachant is constrained, as it will be in a deep geologic repository. Also, once the surface of waste glass is contacted by ground water, the kinetics of establishing solubility control are relatively rapid. The concentrations of leached species reach saturation, or steady-state concentrations, within a few months to a year at 70 to 90/sup 0/C. Thus, reaction kinetics, which were the main subject of earlier leaching mechanisms studies, are now shown to assume much less importance. The dominance of solubility means that the leach rate is, in fact, directly proportional to ground water flow rate. Doubling the flow rate doubles the effective leach rate. This relationship is expected to obtain in most, if not all, repository situations.

This research evaluates the ability of OLI{copyright} equilibrium based software to forecast Savannah River Site HighLevelWaste system impacts from oxalic acid dissolution of Tank 1-15 sludge heels. Without further laboratory and field testing, only the use of oxalic acid can be considered plausible to support sludge heel dissolution on multiple tanks. Using OLI{copyright} and available test results, a dissolution model is constructed and validated. Material and energy balances, coupled with the model, identify potential safety concerns. Overpressurization and overheating are shown to be unlikely. Corrosion induced hydrogen could, however, overwhelm the tank ventilation. While pH adjustment can restore the minimal hydrogen generation, resultant precipitates will notably increase the sludge volume. OLI{copyright} is used to develop a flowsheet such that additional sludge vitrification canisters and other negative system impacts are minimized. Sensitivity analyses are used to assess the processability impacts from variations in the sludge/quantities of acids.

ALPHN calculates the (alpha,n) neutron production rate of a canister of vitrified high-levelwaste. The user supplies the chemical composition of the glass or glass-ceramic and the curies of the alpha-emitting actinides present. The output of the program gives the (alpha,n) neutron production of each actinide in neutrons per second and the total for the canister. The (alpha,n) neutron production rates are source terms only; that is, they are production rates within the glass andmore » do not take into account the shielding effect of the glass. For a given glass composition, the user can calculate up to eight cases simultaneously; these cases are based on the same glass composition but contain different quantities of actinides per canister.« less

The Department of Energy has selected immobilization for disposal in a repository as one approach for disposing of excess plutonium (1). Materials for immobilizing weapons-grade plutonium for repository disposal must meet the ''spent fuel standard'' by providing a radiation field similar to spent fuel (2). Such a radiation field can be provided by incorporating fission products from high-levelwaste into the waste form. Experiments were performed to evaluate the feasibility of incorporating high-levelwaste (HLW) stored at the Idaho Chemical Processing Plant (ICPP) into plutonium dispositioning materials to meet the spent fuel standard. A variety of materials and preparation techniques were evaluated based on prior experience developing waste forms for immobilizing HLW. These included crystalline ceramic compositions prepared by conventional sintering and hot isostatic pressing (HIP), and glass formulations prepared by conventional melting. Because plutonium solubility in silicate melts is limited, glass formulations were intentionally devitrified to partition plutonium into crystalline host phases, thereby allowing increased overall plutonium loading. Samarium, added as a representative rare earth neutron absorber, also tended to partition into the plutonium host phases. Because the crystalline plutonium host phases are chemically more inert, the plutonium is more effectively isolated from the environment, and its attractiveness for proliferation is reduced. In the initial phase of evaluating each material and preparation method, cerium was used as a surrogate for plutonium. For promising materials, additional preparation experiments were performed using plutonium to verify the behavior of cerium as a surrogate. These experiments demonstrated that cerium performed well as a surrogate for plutonium. For the most part, cerium and plutonium partitioned onto the same crystalline phases, and no anomalous changes in oxidation state were observed. The only observed

Highlevel radioactive waste (HLW) is stored in underground storage tanks at the Savannah River Site. The SRS is proceeding with closure of the 22 tanks located in F-Area. Closure consists of removing the bulk of the waste, chemical cleaning, heel removal, stabilizing remaining residuals with tailored grout formulations and severing/sealing external penetrations. A performance assessment is being performed in support of closure of the F-Tank Farm. Initially, the carbon steel construction materials of the highlevelwaste tanks will provide a barrier to the leaching of radionuclides into the soil. However, the carbon steel liners will degrade over time, most likely due to corrosion, and no longer provide a barrier. The tank life estimation in support of the performance assessment has been completed. The estimation considered general and localized corrosion mechanisms of the tank steel exposed to the contamination zone, grouted, and soil conditions. The estimation was completed for Type I, Type III, and Type IV tanks in the F-Tank Farm. The tank life estimation in support of the F-Tank Farm closure performance assessment has been completed. The estimation considered general and localized corrosion mechanisms of the tank steel exposed to the contamination zone, grouted, and soil conditions. The estimation was completed for Type I, Type III, and Type IV tanks in the F-Tank Farm. Consumption of the tank steel encased in grouted conditions was determined to occur either due to carbonation of the concrete leading to low pH conditions, or the chloride-induced de-passivation of the steel leading to accelerated corrosion. A deterministic approach was initially followed to estimate the life of the tank liner in grouted conditions or in soil conditions. The results of this life estimation are shown in Table 1 and Table 2 for grouted and soil conditions respectively. The tank life has been estimated under conservative assumptions of diffusion rates. However, the same process of

The engineering design of disposal of the highlevelwaste (HLW) packages in a geologic repository requires a thermal analysis to provide the temperature history of the packages. Calculated temperatures are used to demonstrate compliance with criteria for waste acceptance into the geologic disposal gallery system and as input to assess the transient thermal characteristics of the vitrified HLW Package. The objective of the work was to evaluate the thermal performance of the supercontainer containing the vitrified HLW in a non-backfilled and unventilated underground disposal gallery. In order to achieve the objective, transient computational models for a geologic vitrified HLW package were developed by using a computational fluid dynamics method, and calculations for the HLW disposal gallery of the current Belgian geological repository reference design were performed. An initial two-dimensional model was used to conduct some parametric sensitivity studies to better understand the geologic system's thermal response. The effect of heat decay, number of co-disposed supercontainers, domain size, humidity, thermal conductivity and thermal emissivity were studied. Later, a more accurate three-dimensional model was developed by considering the conduction-convection cooling mechanism coupled with radiation, and the effect of the number of supercontainers (3, 4 and 8) was studied in more detail, as well as a bounding case with zero heat flux at both ends. The modeling methodology and results of the sensitivity studies will be presented.

The purpose of this calculation is to provide a dose consequence analysis of high-levelwaste (HLW) consisting of plutonium immobilized in vitrified HLW to be handled at the proposed Monitored Geologic Repository at Yucca Mountain for a beyond design basis event (BDBE) under expected conditions using best estimate values for each calculation parameter. In addition to the dose calculation, a plutonium respirable particle size for dose calculation use is derived. The current concept for this waste form is plutonium disks enclosed in cans immobilized in canisters of vitrified HLW (i.e., glass). The plutonium inventory at risk used for this calculation is selected from Plutonium Immobilization Project Input for Yucca Mountain Total Systems Performance Assessment (Shaw 1999). The BDBE examined in this calculation is a nonmechanistic initiating event and the sequence of events that follow to cause a radiological release. This analysis will provide the radiological releases and dose consequences for a postulated BDBE. Results may be considered in other analyses to determine or modify the safety classification and quality assurance level of repository structures, systems, and components. This calculation uses best available technical information because the BDBE frequency is very low (i.e., less than 1.0E-6 events/year) and is not required for License Application for the Monitored Geologic Repository. The results of this calculation will not be used as part of a licensing or design basis.

An evaluation of the optimal filtration conditions was performed based on test data obtained from filtration of a HighLevelWaste Sludge sample from the Hanford tank farms. This evaluation was performed using the anticipated configuration for the Waste Treatment Plant at the Hanford site. Testing was performed to identify the optimal pressure drop and cross flow velocity for filtration at both high and low solids loading. However, this analysis indicates that the actual filtration rate achieved is relatively insensitive to these conditions under anticipated operating conditions. The maximum filter flux was obtained by adjusting the system control valve pressure from 400 to 650 kPa while the filter feed concentration increased from 5 to 20 wt%. However, operating the system with a constant control valve pressure drop of 500 kPa resulted in a less than 1% reduction in the average filter flux. Also note that allowing the control valve pressure to swing as much as +/- 20% resulted in less than a 5% decrease in filter flux.

The ability to effectively mix, sample, certify, and deliver consistent batches of HighLevelWaste (HLW) feed from the Hanford Double Shell Tanks (DST) to the Waste Treatment and Immobilization Plant (WTP) presents a significant mission risk with potential to impact mission length and the quantity of HLW glass produced. DOE’s River Protection Project (RPP) mission modeling and WTP facility modeling assume that individual 3785 cubic meter (1 million gallon) HLW feed tanks are homogenously mixed, representatively sampled, and consistently delivered to the WTP. It has been demonstrated that homogenous mixing of HLW sludge in Hanford DSTs is not likely achievable with the baseline design thereby causing representative sampling and consistent feed delivery to be more difficult. Inconsistent feed to the WTP could cause additional batch to batch operational adjustments that reduces operating efficiency and has the potential to increase the overall mission length. The Hanford mixing and sampling demonstration program will identify DST mixing performance capability, will evaluate representative sampling techniques, and will estimate feed batch consistency. An evaluation of demonstration program results will identify potential mission improvement considerations that will help ensure successful mission completion. This paper will discuss the history, progress, and future activities that will define and mitigate the mission risk.

Job Control Waste (JCW) at the Savannah River Site (SRS) Solid Waste Management Facilities (SWMF) may be disposed of in special containers, and the analysis of these containers requires developing specific analysis methodologies. A method has been developed for the routine assay of prohibited items (liquids, etc.) contained in a 30-gallon drum that is then placed into a 55-gallon drum. Method development consisted of system calibration with a NIST standard at various drum-to-detector distances, method verification with a liquid sample containing a known amount of Pu-238, and modeling the inner container using Ortec Isotopic software. Using this method for measurement of the known standard in the drum-in-drum configuration produced excellent agreement (within 15%) with the known value. Savannah River Site Solid Waste Management also requested analysis of waste contained in large black boxes (commonly 18-feet x 12-feet x 7-feet) stored at the SWMF. These boxes are frequently stored in high background areas and background radiation must be considered for each analysis. A detection limit of less than 150 fissile-gram-equivalents (FGE) of TRU waste is required for the black-box analyses. There is usually excellent agreement for the measurements at different distances and measurement uncertainties of about 50% are obtained at distances of at least twenty feet from the box. This paper discusses the experimental setup, analysis and data evaluation for drum-in-drum and black box waste configurations at SRS.

High-levelwaste (HLW) glass compositions, processing schemes, limits on waste content, and corrosion/dissolution release models are dependent on an accurate knowledge of melting temperatures and thermochemical values. Unfortunately, existing models for predicting these temperatures are empirically-based, depending on extrapolations of experimental information. In addition, present models of leaching behavior of glass waste forms use simplistic assumptions or experimentally measured values obtained under non-realistic conditions. There is thus a critical need for both more accurate and more widely applicable models for HLW glass behavior, which this project addressed. Significant progress was made in this project on modeling HLW glass. Borosilicate glass was accurately represented along with the additional important components that contain iron, lithium, potassium, magnesium, and calcium. The formation of crystalline inclusions in the glass, an issue in Hanford HLW formulations, was modeled and shown to be predictive. Thus the results of this work have already demonstrated practical benefits with the ability to map compositional regions where crystalline material forms, and therefore avoid that detrimental effect. With regard to a fundamental understanding, added insights on the behavior of the components of glass have been obtained, including the potential formation of molecular clusters. The EMSP project had very significant effects beyond the confines of Environmental Management. The models developed for glass have been used to solve a very costly problem in the corrosion of refractories for glass production. The effort resulted in another laboratory, Sandia National Laboratories-Livermore, to become conversant in the techniques and to apply those through a DOE Office of Industrial Technologies project joint with PPG Industries. The glass industry as a whole is now cognizant of these capabilities, and there is a Glass Manufacturer's Research Institute proposal

In accordance with the Nuclear Waste Policy Amendments Act of 1987, Yucca Mountain was designated as the site to be investigated as a potential repository for the disposal of high-level radioactive waste. The Yucca Mountain site is an undeveloped area located on the southwestern edge of the Nevada Test Site (NTS), about 100 miles northwest of Las Vegas. The site currently lacks rail service or an existing right-of-way. If the Yucca Mountain site is found suitable for the repository, rail service is desirable to the Office of Civilian Waste Management (OCRWM) Program because of the potential of rail transportation to reduce costs and to reduce the number of shipments relative to highway transportation. A Preliminary Rail Access Study evaluated 13 potential rail spur options. Alternative routes within the major options were also developed. Each of these options was then evaluated for potential land use conflicts and access to regional rail carriers. Three potential routes having few land use conflicts and having access to regional carriers were recommended for further investigation. Figure 1-1 shows these three routes. The Jean route is estimated to be about 120 miles long, the Carlin route to be about 365 miles long, and Caliente route to be about 365 miles long. The remaining ten routes continue to be monitored and should any of the present conflicts change, a re-evaluation of that route will be made. Complete details of the evaluation of the 13 routes can be found in the previous study. The DOE has not identified any preferred route and recognizes that the transportation issues need a full and open treatment under the National Environmental Policy Act. The issue of transportation will be included in public hearings to support development of the Environmental Impact Statement (EIS) proceedings for either the Monitored Retrievable Storage Facility or the Yucca Mountain Project or both.

This document addresses the preferred combination of design and operational configurations to provide heat removal from high-levelwaste tanks during Phase 1 waste feed delivery to prevent the waste temperature from exceeding tank safety requirement limits. An interim decision for the preferred method to remove the heat from the high-levelwaste tanks during waste feed delivery operations is presented herein.

Kineococcus radiotolerans SRS30216 was isolated from a high-level radioactive environment at the Savannah River Site (SRS) and exhibits γ-radiation resistance approaching that of Deinococcus radiodurans. The genome was sequenced by the U.S. Department of Energy's Joint Genome Institute which suggested the existence of three replicons, a 4.76 Mb linear chromosome, a 0.18 Mb linear plasmid, and a 12.92 Kb circular plasmid. Southern hybridization confirmed that the chromosome is linear. The K. radiotolerans genome sequence was examined to learn about the physiology of the organism with regard to ionizing radiation resistance, the potential for bioremediation of nuclear waste, and the dimorphic life cycle. K. radiotolerans may have a unique genetic toolbox for radiation protection as it lacks many of the genes known to confer radiation resistance in D. radiodurans. Additionally, genes involved in the detoxification of reactive oxygen species and the excision repair pathway are overrepresented. K. radiotolerans appears to lack degradation pathways for pervasive soil and groundwater pollutants. However, it can respire on two organic acids found in SRShigh-level nuclear waste, formate and oxalate, which promote the survival of cells during prolonged periods of starvation. The dimorphic life cycle involves the production of motile zoospores. The flagellar biosynthesis genes are located on a motility island, though its regulation could not be fully discerned. These results highlight the remarkable ability of K radiotolerans to withstand environmental extremes and suggest that in situ bioremediation of organic complexants from highlevel radioactive waste may be feasible. PMID:19057647

This report evaluates the effects of feed composition on the Phase I high-levelwaste immobilization process and interim storage facility requirements for the high-levelwaste glass.Several different Phase I staging (retrieval, blending, and pretreatment) scenarios were used to generate example feed compositions for glass formulations, testing, and glass sensitivity analysis. Glass models and data form laboratory glass studies were used to estimate achievable waste loading and corresponding glass volumes for various Phase I feeds. Key issues related to feed process ability, feed composition, uncertainty, and immobilization process technology are identified for future consideration in other tank waste disposal program activities.

High-levelwaste will be placed in sealed waste packages inside a shielded closure cell. The Idaho National Laboratory (INL) has designed a system for closing the waste packages including all cell interior equipment and support systems. This paper discusses the material handling aspects of the equipment used and operations that will take place as part of the waste package closure operations. Prior to construction, the cell and support system will be assembled in a full-scale mockup at INL.

The Nuclear Waste Policy Act of 1982 mandated that all high-levelwaste (HLW) be sent to a federal geologic repository for permanent disposal. DOE published the Environmental Assessment in 1982 which identified borosilicate glass as the chosen HLW form.{sup 1} In 1985 the Department of Energy instituted a Waste Acceptance Process to assure that DWPF glass waste forms would be acceptable to such a repository. This assurance was important since production of waste forms will precede repository construction and licensing. As part of this Waste Acceptance Process, the DOE Office of Civilian Radioactive Waste Management (RW) formed the Waste Acceptance Committee (WAC). The WAC included representatives from the candidate repository sites, the waste producing sites and DOE. The WAC was responsible for developing the Waste Acceptance Preliminary Specifications (WAPS) which defined the requirements the waste forms must meet to be compatible with the candidate repository geologies.

The Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS) will be the United States` first facility to process HighLevel radioactive Waste (HLW) into a borosilicate glass matrix. The removal of aromatic precipitates by hydrolysis, evaporation, liquid-liquid extraction and decantation will be a key step in the processing of the HLW. This step, titled the Precipitate Hydrolysis Process, has been demonstrated by the Savannah River Technology Center with the Precipitate Hydrolysis Experimental Facility (PHEF). The mission of the PHEF is to demonstrate processing of simulated highlevel radioactive waste which contains tetraphenylborate precipitates and nitrite. Aqueous washing or nitrite destruction is used to reduce nitrite. Formic acid with a copper catalyst is used to hydrolyze tetraphenylborate (TPB). The primary offgases are benzene, carbon dioxide, nitrous oxide, and nitric oxide. Hydrolysis of TPB in the presence of nitrite results in the production of polycyclic aromatics and aromatic amines (referred as high boiling organics) such as biphenyl, diphenylamine, terphenyls etc. The decanter separates the organic (benzene) and aqueous phase, but the high boiling organic separation is difficult. This paper focuses on the evaluation of the operating strategies, including steam stripping, to maximize the removal of the high boiling organics from the aqueous stream. Two areas were investigated, (1) a stream stripping comparison of the late wash flowsheet to the HAN flowsheet and (2) the extraction performance of the original decanter to the new decanter. The focus of both studies was to minimize the high boiling organic content of the Precipitate Hydrolysis Aqueous (PHA) product in order to minimize downstream impacts caused by organic deposition.

The U.S. Department of Energy (DOE) is building a Tank Waste Treatment and Immobilization Plant (WTP) at the Hanford Site in Washington to remediate 55 million gallons of radioactive waste that is being temporarily stored in 177 underground tanks. Efforts are being made to increase the loading of Hanford tank wastes in glass while meeting melter lifetime expectancies and process, regulatory, and product quality requirements. This road map guides the research and development for formulation and processing of crystaltolerant glasses, identifying near- and long-term activities that need to be completed over the period from 2014 to 2019. The primary objective is to maximize waste loading for Hanford waste glasses without jeopardizing melter operation by crystal accumulation in the melter or melter discharge riser. The potential applicability to the Savannah River Site (SRS) Defense Waste Processing Facility (DWPF) will also be addressed in this road map. The planned research described in this road map is motivated by the potential for substantial economic benefits (significant reductions in glass volumes) that will be realized if the current constraints (T1% for WTP and TL for DWPF) are approached in an appropriate and technically defensible manner for defense waste and current melter designs. The basis of this alternative approach is an empirical model predicting the crystal accumulation in the WTP glass discharge riser and melter bottom as a function of glass composition, time, and temperature. When coupled with an associated operating limit (e.g., the maximum tolerable thickness of an accumulated layer of crystals), this model could then be integrated into the process control algorithms to formulate crystal-tolerant high-levelwaste (HLW) glasses targeting high waste loadings while still meeting process related limits and melter lifetime expectancies. The modeling effort will be an iterative process, where model form and a broader range of conditions, e.g., glass

Many types of ceramics have been put forward for immobilisation of high-levelwaste (HLW) from reprocessing of nuclear power plant fuel or weapons production. After describing some historical aspects of waste form research, the essential features of the chemical design and processing of these different ceramic types will be discussed briefly. Given acceptable laboratory and long-term predicted performance based on appropriately rigorous chemical design, the important processing parameters are mostly waste loading, waste throughput, footprint, offgas control/minimization, and the need for secondary waste treatment. It is concluded that the 'problem of high-level nuclear waste' is largely solved from a technical point of view, within the current regulatory framework, and that the main remaining question is which technical disposition method is optimum for a given waste. (author)

Forty three of the HighLevelWaste (HLW) tanks at the Savannah River Site (SRS) have internal structures that hinder removal of the last approximately five thousand gallons of waste sludge solely by mechanical means. Chemical cleaning can be utilized to dissolve the sludge heel with oxalic acid (OA) and pump the material to a separate waste tank in preparation for final disposition. This dissolved sludge material is pH adjusted downstream of the dissolution process, precipitating the sludge components along with sodium oxalate solids. The large quantities of sodium oxalate and other metal oxalates formed impact downstream processes by requiring additional washing during sludge batch preparation and increase the amount of material that must be processed in the tank farm evaporator systems and the Saltstone Processing Facility. Enhanced Chemical Cleaning (ECC) was identified as a potential method for greatly reducing the impact of oxalate additions to the SRS Tank Farms without adding additional components to the waste that would extend processing or increase waste form volumes. In support of Savannah River Site (SRS) tank closure efforts, the Savannah River National Laboratory (SRNL) conducted Real Waste Testing (RWT) to evaluate an alternative to the baseline 8 wt. % OA chemical cleaning technology for tank sludge heel removal. The baseline OA technology results in the addition of significant volumes of oxalate salts to the SRS tank farm and there is insufficient space to accommodate the neutralized streams resulting from the treatment of the multiple remaining waste tanks requiring closure. ECC is a promising alternative to bulk OA cleaning, which utilizes a more dilute OA (nominally 2 wt. % at a pH of around 2) and an oxalate destruction technology. The technology is being adapted by AREVA from their decontamination technology for Nuclear Power Plant secondary side scale removal. This report contains results from the SRNL small scale testing of the ECC process

Large areas of the deep seabed warrant assessment as potential disposal sites for high-level radioactive waste because: (1) they are far from seismically and tectonically active lithospheric plate boundaries; (2) they are far from active or young volcanos; (3) they contain thick layers of very uniform fine-grained clays; (4) they are devoid of natural resources likely to be exploited in the forseeable future; (5) the geologic and oceanographic processes governing the deposition of sediments in such areas are well understood, and are remarkably insensitive to past oceanographic and climatic changes; and (6) sedmentary records of tens of millions of years of slow, uninterrupted deposition of fine grained clay support predictions of the future stability of such sites. Data accumulated to date on the permeability, ion-retardation properties, and mechanical strength of pelagic clay sediments indicate that they can act as a primary barrier to the escape of buried nuclides. Work in progress should determine within the current decade whether subseabed disposal is environmentally acceptable and technically feasible, as well as address the legal, political and social issues raised by this new concept.

The Defense Waste Processing Facility at the Savannah River Site treats and immobilizes HighLevelWaste into a durable borosilicate glass for safe, permanent storage. The HighLevelWaste program significantly reduces environmental risks associated with the storage of radioactive waste from legacy efforts to separate fissionable nuclear material from irradiated targets and fuels. In an effort to support the disposition of radioactive waste and accelerate tank closure at the Savannah River Site, the Defense Waste Processing Facility recently implemented facility and flowsheet modifications to improve production by 25%. These improvements, while low in cost, translated to record facility production in fiscal years 2011 and 2012. In addition, significant progress has been accomplished on longer term projects aimed at simplifying and expanding the flexibility of the existing flowsheet in order to accommodate future processing needs and goals. (authors)

Sampling residual waste after tank cleaning and analysis for analytes of concern to support closure and cleaning targets of large underground tanks used for storage of legacy highlevel radioactive waste (HLW) at Department of Energy (DOE) sites has been underway since about 1995. The DOE Tanks Focus Area (TFA) has been working with DOE tank sites to develop new sampling plans, and sampling methods for assessment of residual waste inventories. This paper discusses regulatory analytes of concern, sampling plans, and sampling methods that support closure and cleaning target activities for large storage tanks at the Hanford Site, the Savannah River Site (SRS), the Idaho National Engineering and Environmental Laboratory (INEEL), and the West Valley Demonstration Project (WVDP).

Sampling residual waste after tank cleaning and analysis for analytes of concern to support closure and cleaning targets of large underground tanks used for storage of legacy highlevel radioactive waste (HLW) at Department of Energy (DOE) sites has been underway since about 1995. The DOE Tanks Focus Area (TFA) has been working with DOE tank sites to develop new sampling plans, and sampling methods for assessment of residual waste inventories. This paper discusses regulatory analytes of concern, sampling plans, and sampling methods that support closure and cleaning target activities for large storage tanks at the Hanford Site, the Savannah River Site (SRS), the Idaho National Engineering and Environmental Laboratory (INEEL), and the West Valley Demonstration Project (WVDP).

The U.S. Department of Energy (DOE) has embarked upon a course to acquire Hanford Site tank waste treatment and immobilization services using privatized facilities (RL 1996a). This plan contains a two-phased approach. Phase I is a proof-of-principle/connnercial demonstration- scale effort and Phase II is a fiill-scale production effort. In accordance with the planned approach, interim storage and disposal of various products from privatized facilities are to be DOE fumished. The high-levelwaste (BLW) interim storage options, or alternative architectures, were identified and evaluated to provide the framework from which to select the most viable method of Phase I BLW interim storage (Calmus 1996). This evaluation, hereafter referred to as the Alternative Architecture Evaluation, was performed to established performance and risk criteria (technical merit, cost, schedule, etc.). Based on evaluation results, preliminary architectures and path forward reconunendations were provided for consideration in the architecture decision- maldng process. The decision-making process used for selection of a Phase I solidified BLW interim storage architecture was conducted in accordance with an approved Decision Plan (see the attachment). This decision process was based on TSEP-07,Decision Management Procedure (WHC 1995). The established decision process entailed a Decision Board, consisting of Westinghouse Hanford Company (VY`HC) management staff, and included appointment of a VTHC Decision Maker. The Alternative Architecture Evaluation results and preliminary recommendations were presented to the Decision Board members for their consideration in the decision-making process. The Alternative Architecture Evaluation was prepared and issued before issuance of @C-IP- 123 1, Alternatives Generation and Analysis Procedure (WI-IC 1996a), but was deemed by the Board to fully meet the intent of WHC-IP-1231. The Decision Board members concurred with the bulk of the Alternative Architecture

High-levelwaste (HLW) throughput (i.e., the amount of waste processed per unit time) is a function of two critical parameters: waste loading (WL) and melt rate. For the Waste Treatment and Immobilization Plant (WTP) at the Hanford Site and the Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS), increasing HLW throughput would significantly reduce the overall mission life cycle costs for the Department of Energy (DOE). The objective of this task is to develop data, assess property models, and refine or develop the necessary models to support increased WL of HLW at SRS. It is a continuation of the studies initiated in FY07, but is under the specific guidance of a Task Change Request (TCR)/Work Authorization received from DOE headquarters (Project Number RV071301). Using the data generated in FY07, FY08 and historical data, two test matrices (60 glasses total) were developed at the Savannah River National Laboratory (SRNL) in order to generate data in broader compositional regions. These glasses were fabricated and characterized using chemical composition analysis, X-ray Diffraction (XRD), viscosity, liquidus temperature (TL) measurement and durability as defined by the Product Consistency Test (PCT). The results of this study are summarized below: (1) In general, the current durability model predicts the durabilities of higher waste loading glasses quite well. A few of the glasses exhibited poorer durability than predicted. (2) Some of the glasses exhibited anomalous behavior with respect to durability (normalized leachate for boron (NL [B])). The quenched samples of FY09EM21-02, -07 and -21 contained no nepheline or other wasteform affecting crystals, but have unacceptable NL [B] values (> 10 g/L). The ccc sample of FY09EM21-07 has a NL [B] value that is more than one half the value of the quenched sample. These glasses also have lower concentrations of Al{sub 2}O{sub 3} and SiO{sub 2}. (3) Five of the ccc samples (EM-13, -14, -15, -29 and

This road map guides the research and development for formulation and processing of crystal-tolerant glasses, identifying near- and long-term activities that need to be completed over the period from 2014 to 2019. The primary objective is to maximize waste loading for Hanford waste glasses without jeopardizing melter operation by crystal accumulation in the melter or melter discharge riser. The potential applicability to the Savannah River Site (SRS) Defense Waste Processing Facility (DWPF) is also addressed in this road map.

Three Savannah River Laboratory reference high-levelwaste canisters were subjected to impact tests at the Pacific Northwest Laboratory in Richland, Washington, in June 1983. The purpose of the test was to determine the integrity of the canister, nozzle, and final closure weld and to assess the effects of impacts on the glass. Two of the canisters were fabricated from 304L stainless steel and the third canister from titanium. The titanium canister was subjected to two drops. The first drop was vertical from 9.14 m onto an unyielding surface with the bottom corner of the canister receiving the impact. No failure occurred during this drop. The second drop was vertical from 9.14 m onto an unyielding surface with the corner of the fill nozzle receiving the impact. A large breach in the canister occurred in the region where the fill nozzle joins the dished head. The first stainless steel canister was dropped with the corner of the fill nozzle receiving the impact. The canister showed significant strain with no rupturing in the region where the fill nozzle joins the dished head. The second canister was dropped with the bottom corner receiving the impact and also, dropped horizontally onto an unyielding vertical solid steel cylinder in a puncture test. The bottom drop did not damage the weld and the puncture test did not rupture the canister body. The glass particles in the damaged zone of these canisters were sampled and analyzed for particle size. A comparison was made with control canister in which no impact had occurred. The particle size distribution for the control canisters and the zones of damaged glass were determined down to 1.5 ..mu..m. The quantity of glass fines, smaller than 10 ..mu..m, which must be determined for transportation safety studies, was found to be the largest in the bottom-damaged zone. The total amount of fines smaller than 10 ..mu..m after impact was less than 0.01 wt % of the total amount of glass in the canister.

This report predicts 30 year receipts of solid transuranic (TRU) wastes from eventual plutonium facility deactivation and cleanup, and combines them with the existing TRU waste holdings to provide a technical and quantitative basis for interim and long range TRU waste management planning. The current TRU waste holdings have been characterized based on data from the Computerized Radioactive Waste Burial Records Analysis (COBRA) system. Six TRU waste disposition categories have been identified for existing TRU waste as shown in Table 1. An additional category has been quantified that includes projected waste volumes from the Decontamination and Decommissioning (D&D) of TRU waste generating facilities. These projections are based on COBRA data from D&D of the original plutonium finishing facilities in F and H Areas that were replaced in the 1970`s and 80`s.

Twenty five glasses were formulated. They were batched from HLW AZ-101 simulant or raw chemicals and melted and tested with a series of tests to elucidate the effect of spinel-forming components (Ni, Fe, Cr, Mn, and Zn), Al, and noble metals (Rh2O3 and RuO2) on the accumulation rate of spinel crystals in the glass discharge riser of the high-levelwaste (HLW) melter. In addition, the processing properties of glasses, such as the viscosity and TL, were measured as a function of temperature and composition. Furthermore, the settling of spinel crystals in transparent low-viscosity fluids was studied at room temperature to access the shape factor and hindered settling coefficient of spinel crystals in the Stokes equation. The experimental results suggest that Ni is the most troublesome component of all the studied spinel-forming components producing settling layers of up to 10.5 mm in just 20 days in Ni-rich glasses if noble metals or a higher concentration of Fe was not introduced in the glass. The layer of this thickness can potentially plug the bottom of the riser, preventing glass from being discharged from the melter. The noble metals, Fe, and Al were the components that significantly slowed down or stopped the accumulation of spinel at the bottom. Particles of Rh2O3 and RuO2, hematite and nepheline, acted as nucleation sites significantly increasing the number of crystals and therefore decreasing the average crystal size. The settling rate of ≤10-μm crystal size around the settling velocity of crystals was too low to produce thick layers. The experimental data for the thickness of settled layers in the glasses prepared from AZ-101 simulant were used to build a linear empirical model that can predict crystal accumulation in the riser of the melter as a function of concentration of spinel-forming components in glass. The developed model predicts the thicknesses of accumulated layers quite well, R2 = 0.985, and can be become an efficient tool for the formulation

The High-LevelWaste Data Base is a menu-driven PC data base developed as part of OCRWM's technical data base on the characteristics of potential repository wastes, which also includes spent fuel and other materials. This programmer's guide completes the documentation for the High-LevelWaste Data Base, the user's guide having been published previously. 3 figs.

The Hanford Tank Operations Contractor (TOC) and the Hanford Waste Treatment and Immobilization Plant (WTP) contractor are both engaged in demonstrating mixing, sampling, and transfer system capability using simulated Hanford High-LevelWaste (HL W) formulations. This work represents one of the remaining technical issues with the high-levelwaste treatment mission at Hanford. The TOC must demonstrate the ability to adequately mix and sample high-levelwaste feed to meet the WTP Waste Acceptance Criteria and Data Quality Objectives. The sampling method employed must support both TOC and WTP requirements. To facilitate information transfer between the two facilities the mixing and sampling demonstrations are led by the One System Integrated Project Team. The One System team, Waste Feed Delivery Mixing and Sampling Program, has developed a full scale sampling loop to demonstrate sampler capability. This paper discusses the full scale sampling loops ability to meet precision and accuracy requirements, including lessons learned during testing. Results of the testing showed that the Isolok(R) sampler chosen for implementation provides precise, repeatable results. The Isolok(R) sampler accuracy as tested did not meet test success criteria. Review of test data and the test platform following testing by a sampling expert identified several issues regarding the sampler used to provide reference material used to judge the Isolok's accuracy. Recommendations were made to obtain new data to evaluate the sampler's accuracy utilizing a reference sampler that follows good sampling protocol.

The general thermal hydraulics program GOTH-SNF was used to predict the thermal response of the waste in tanks 241-AY-102 and 241-AZ-102 when mixed by two 300 horsepower mixer pumps. This mixing was defined in terms of a specific waste retrieval scenario. Both dome and annulus ventilation system flow are necessary to maintain the waste within temperature control limits during the mixing operation and later during the sludge-settling portion of the scenario are defined.

Radioactive waste containing low concentrations of fissile isotopes is stored in underground storage tanks on the Hanford Site in Washington State. The goal of criticality safety is to ensure that this waste remains subcritical into the indefinite future without supervision. A large ratio of solids to plutonium provides an effective way of ensuring a low plutonium concentration. Since the first waste discharge, a program of audits and appraisals has ensured that operations are conducted according to limits and controls applied to them. In addition, a program of surveillance and characterization maintains watch over waste after discharge.

The objective of the proposed effort is to expand the development of solution models of complex waste glass systems that are predictive with regard to composition, phase separation, chemical activity, and volatility. The effort will also yield thermodynamic values for waste compo...

The objective of the proposed effort was to expand the development of solution models of complex waste glass systems that are predictive with regard to composition, phase separation, chemical activity, and volatility. The effort was to yield thermodynamic values for waste compone...

Described are problems and techniques for safe disposal of radioactive waste. Degrees of radioactivity, temporary storage, and long-term permanent storage are discussed. Included are diagrams of estimated waste volumes to the year 2000 and of an artist's conception of a permanent underground disposal facility. (SL)

A number of waste components in US defense highlevel radioactive wastes (HLW) have proven challenging for current Joule heated ceramic melter (JCHM) operations and have limited the ability to increase waste loadings beyond already realized levels. Many of these ''troublesome'' waste species cause crystallization in the glass melt that can negatively impact product quality or have a deleterious effect on melter processing. Thus, recent efforts at US Department of Energy laboratories have focused on understanding crystallization behavior within HLW glass melts and investigating approaches to mitigate the impacts of crystallization so that increases in waste loading can be realized.more » Advanced glass formulations have been developed to highlight the unique benefits of next-generation melter technologies such as the Cold Crucible Induction Melter (CCIM). Crystal-tolerant HLW glasses have been investigated to allow sparingly soluble components such as chromium to crystallize in the melter but pass out of the melter before accumulating. The Hanford site AZ-101 tank waste composition represents a waste group that is waste loading limited primarily due to high concentrations of Fe2O3 (with higher Al2O3). Systematic glass formulation development utilizing slightly higher process temperatures and higher tolerance to spinel crystals demonstrated that an increase in waste loading of more than 20% could be achieved for this waste composition, and by extension higher loadings for wastes in the same group.« less

A number of waste components in US defense highlevel radioactive wastes (HLW) have proven challenging for current Joule heated ceramic melter (JCHM) operations and have limited the ability to increase waste loadings beyond already realized levels. Many of these ''troublesome'' waste species cause crystallization in the glass melt that can negatively impact product quality or have a deleterious effect on melter processing. Thus, recent efforts at US Department of Energy laboratories have focused on understanding crystallization behavior within HLW glass melts and investigating approaches to mitigate the impacts of crystallization so that increases in waste loading can be realized. Advanced glass formulations have been developed to highlight the unique benefits of next-generation melter technologies such as the Cold Crucible Induction Melter (CCIM). Crystal-tolerant HLW glasses have been investigated to allow sparingly soluble components such as chromium to crystallize in the melter but pass out of the melter before accumulating. The Hanford site AZ-101 tank waste composition represents a waste group that is waste loading limited primarily due to high concentrations of Fe2O3 (with higher Al2O3). Systematic glass formulation development utilizing slightly higher process temperatures and higher tolerance to spinel crystals demonstrated that an increase in waste loading of more than 20% could be achieved for this waste composition, and by extension higher loadings for wastes in the same group.

The US Department of Energy, Hanford Site nuclear reservation, located in Southeastern Washington State, is currently home to 61 Mgal of radioactive waste stored in 177 large underground storage tanks. As an intermediate waste volume reduction, the 242-A Evaporator/Crystallizer processes waste solutions from most of the operating laboratories and plants on the Hanford Site. The waste solutions are concentrated in the Evaporator/Crystallizer to a slurry of liquid and crystallized salts. This concentrated slurry is returned to Hanford Site waste tanks at a significantly reduced volume. The Washington State Department of Ecology Dangerous Waste Regulations, WAC 173-393 require that a tank system integrity assessment be completed and maintained on file at the facility for all dangerous waste tank systems. This corrosion evaluation was performed in support of the 242-A Evaporator/Crystallizer Tank System Integrity Assessment Report. This corrosion evaluation provided a comprehensive compatibility study of the component materials and corrosive environments. Materials used for the Evaporator components and piping include austenitic stainless steels (SS) (primarily ASTM A240, Type 304L) and low alloy carbon steels (CS) (primarily ASTM A53 and A106) with polymeric or asbestos gaskets at flanged connections. Building structure and secondary containment is made from ACI 301-72 Structural Concrete for Buildings and coated with a chemically resistant acrylic coating system.

High-level nuclear waste produced from fuel reprocessing operations at the Savannah River Site (SRS) requires pretreatment to remove Cs-137, Sr-90, and alpha-emitting radionuclides (i.e., actinides) prior to disposal onsite as low level waste. Separation processes planned at SRS include caustic side solvent extraction, for Cs-137 removal, and sorption of Sr-90 and alpha-emitting radionuclides onto monosodium titanate (MST). The predominant alpha-emitting radionuclides in the highly alkaline waste solutions include plutonium isotopes Pu-238, Pu-239, and Pu-240. This paper describes recent results from the development of an improved titanate material that exhibits increased removal kinetics and effective capacity for Sr-90 and alpha-emitting radionuclides compared to the baseline MST material.

The Savannah River Technology Center (SRTC) has been funded through the DOE Office of Technology Development (DOE-OTD) to investigate high-temperature vitrification technologies for the treatment of diverse low-level and mixed wastes. High temperature vitrification is a likely candidate for processing heterogeneous solid wastes containing low levels of activity. Many SRSwastes fit into this category. Plasma torch technology is one high temperature vitrification method. A trial demonstration of plasma torch processing is being performed at the Georgia Institute of Technology on surrogate SRSwastes. This effort is in cooperation with the Engineering Research and Development Association of Georgia Universities (ERDA) program. The results of phase 1 of these plasma torch trials will be presented.

This study evaluates the effect of feed composition on the performance of the high-level vitrification process. It is assumed in this study that the tank wastes are retrieved and blended by tank farms, producing 12 different blends from the single-shell tank farms, two blends of double-shell tank waste, and a separately defined all-tank blend. This blending scenario was chosen only for evaluating the impact of composition on the volume of high- levelwaste glass produced. Special glass compositions were formulated for each waste blend based on glass property models and the properties of similar glasses. These glasses were formulated to meet the applicable viscosity, electrical conductivity, and liquidus temperature constraints for the identified candidate melters. Candidate melters in this study include the low-temperature stirred melter, which operates at 1050{degrees}C; the reference Hanford Waste Vitrification Plant liquid-fed ceramic melter, which operates at 1150{degrees}C; and the high-temperature, joule-heated melter and the cold-crucible melter, which operate over a temperature range of 1150{degrees}C to 1400{degrees}C. In the most conservative case, it is estimated that 61,000 MT of glass will be produced if the Site`s high-levelwastes are retrieved by tank farms and processed in the reference joule-heated melter. If an all-tank blend was processed under the same conditions, the reference melter would produce 21,250 MT of glass. If cross-tank blending were used, it is anticipated that $2.0 billion could be saved in repository disposal costs (based on an average disposal cost of $217,000 per canister) by blending the S, SX, B, and T Tank Farm wastes with other wastes prior to vitrification. General blending among all the tank farms is expected to produce great potential benefit.

Two options for Environmental Impact Statement No action alternatives describe operation of the radioactive liquid waste facilities at the Idaho Chemical Processing Plant at the Idaho National Engineering and Environmental Laboratory. The first alternative describes continued operation of all facilities as planned and budgeted through 2020. Institutional control for 100 years would follow shutdown of operational facilities. Alternatively, the facilities would be shut down in an orderly fashion without completing planned activities. The facilities and associated operations are described. Remaining sodium bearing liquid waste will be converted to solid calcine in the New Waste Calcining Facility (NWCF) or will be left in the waste tanks. The calcine solids will be stored in the existing Calcine Solids Storage Facilities (CSSF). Regulatory and cost impacts are discussed.

Moral philosophy applied to nuclear waste disposal can be linked to paradigmatic science. Simple thermodynamic principles tell us something about rightness or wrongness of our action. Ethical judgement can be orientated towards the chemical compatibility between waste container and geological repository. A container-repository system as close as possible to thermodynamic equilibrium is ethically acceptable. It aims at unlimited stability, similar to the stability of natural metal deposits within the Earth’s crust. The practicability of the guideline can be demonstrated.

A number of waste components in US defense highlevel radioactive wastes (HLW) have proven challenging for current Joule heated ceramic melter (JHCM) operations and have limited the ability to increase waste loadings beyond already realized levels. Many of these "troublesome" waste species cause crystallization in the glass melt that can negatively impact product quality or have a deleterious effect on melter processing. Recent efforts at US Department of Energy laboratories have focused on understanding crystallization behavior within HLW glass melts and investigating approached to mitigate the impacts of crystallization so that increases in waste loading can be realized. Advanced glass formulations have been developed to highlight the unique benefits of next-generation melter technologies such as the Cold Crucible Induction Melter (CCIM). Crystal-tolerant HLW glasses have been investigated to allow sparingly soluble components such as chromium to crystallize in the melter but pass out of the melter before accumulating.

Boehmite is present in significant quantities in several of the Hanford waste tanks. It has been proposed that the boehmite will be dissolved through caustic leaching in the Hanford Waste Treatment Plant currently under construction. Therefore, it is important to fully understand the nature of this dissolution so that the process can be deployed. This research determined the impact of primary control parameters on the boehmite dissolution rate. The impact of aluminate ion on the dissolution kinetics was determined. In addition, other parameters that impact boehmite dissolution, such as free hydroxide concentration and reaction temperature, were also assessed and used to develop a semi-empirical model of the boehmite dissolution process. The understanding derived from this work will be used as the basis to evaluate and improve the planned performance of the Hanford Waste Treatment plant. This work is the first in a series of programs aimed at demonstrating the Waste Treatment Plant dissolution process. This work will be used to develop a simulant of the boehmite-containing Hanford waste. That simulant will then be used in laboratory- and pilot-scale testing to demonstrate the Waste Treatment Plant pretreatment process in an integrated fashion.

In this dissertation different aspects of corrosion and electrochemistry of copper, candidate canister material in Scandinavian high-level nuclear waste disposal program, including the thermodynamics and kinetics of the reactions that are predicted to occur in the practical system have been studied. A comprehensive thermodynamic study of copper in contact with granitic groundwater of the type and composition that is expected in the Forsmark repository in Sweden has been performed. Our primary objective was to ascertain whether copper would exist in the thermodynamically immune state in the repository, in which case corrosion could not occur and the issue of corrosion in the assessment of the storage technology would be moot. In spite of the fact that metallic copper has been found to exist for geological times in granitic geological formations, copper is well-known to be activated from the immune state to corrode by specific species that may exist in the environment. The principal activator of copper is known to be sulfur in its various forms, including sulfide (H2S, HS-, S2-), polysulfide (H2Sx, HSx -, Sx 2-), poly sulfur thiosulfate ( SxO3 2-), and polythionates (SxO6 2-). A comprehensive study of this aspect of copper chemistry has never been reported, and yet an understanding of this issue is vital for assessing whether copper is a suitable material for fabricating canisters for the disposal of HLNW. Our study identifies and explores those species that activate copper; these species include sulfur-containing entities as well as other, non-sulfur species that may be present in the repository. The effects of temperature, solution pH, and hydrogen pressure on the kinetics of the hydrogen electrode reaction (HER) on copper in borate buffer solution have been studied by means of steady-state polarization measurements, including electrochemical impedance spectroscopy (EIS). In order to obtain electrokinetic parameters, such as the exchange current density and the

This report assesses the tank farm`s highlevelwaste supernate to determine any potential impacts on waste certification for the E-Area vaults (EAV). The Waste Acceptance Criteria procedure (i.e., WAC 3.10 of the 1S manual) imposes administrative controls on radioactive material in waste packages sent to the EAV, specifically on six fission products. Waste tank supernates contain various fission products, so any waste package containing material contaminated with supernate will contain these radioactive isotopes. This report develops the process knowledge basis for characterizing the supernate composition for these isotopes, so that appropriate controls can be implemented to ensure that the EAV WAC is met. Six fission products are listed in the SRS 1S Manual WAC 3.10: Se-79, which decays to bromine; Sr-90, which decays to niobium; Tc-99, which decays to ruthenium; Sn-126, which decays to tellurium; I-129, which decays to xenon; and Cs-137, which decays to barium.

Highly radioactive wastes stored at Hanford and Savannah River DOE sites have unresolved questions relating to safety of the stored waste, as well as needs for safe, effective, and efficient waste processing to minimize the volume of high-levelwaste (HLW) streams for disposal. ...

Federal regulatory criteria for geologic disposal of high-levelwaste are under development. Also, interim performance specifications for high-levelwaste forms in geologic isolation are being developed within the Federal program responsible for repository selection and operation. Two high-levelwaste forms, borosilicate glass and crystalline ceramic, have been selected as candidate immobilization forms for the Defense Waste Processing Facility (DWPF) which is to immobilize high-levelwastes at the Savannah River Plant (SRP). An assessment of how these two waste forms conform with the proposed regulatory criteria and repository specifications was performed. Both forms were determined to be in conformance with postulated rules for radionuclide releases and radiation exposures throughout the entire waste disposal system, as well as with proposed repository operation requirements.

Information concerning the condition of the high-levelwaste tanks at the Western New York State Nuclear Service center near West Valley, New York is presented. This information is to be used in evaluating the safety of continued storage and in the development of alternatives for final disposition of the high-levelwaste.

A number of waste components in US defense highlevel radioactive wastes (HLW) have proven challenging for current Joule heated ceramic melter (JHCM) operations and have limited the ability to increase waste loadings beyond already realized levels. Many of these “troublesome" waste species cause crystallization in the glass melt that can negatively impact product quality or have a deleterious effect on melter processing. Recent efforts at US Department of Energy laboratories have focused on understanding crystallization behavior within HLW glass melts and investigating approaches to mitigate the impacts of crystallization so that increases in waste loading can be realized. Advanced glass formulations have been developed to highlight the unique benefits of next-generation melter technologies such as the Cold Crucible Induction Melter (CCIM). Crystal-tolerant HLW glasses have been investigated to allow sparingly soluble components such as chromium to crystallize in the melter but pass out of the melter before accumulating.The Hanford site AZ-101 tank waste composition represents a waste group that is waste loading limited primarily due to high concentrations of Fe2O3 (also with high Al2O3 concentrations). Systematic glass formulation development utilizing slightly higher process temperatures and higher tolerance to spinel crystals demonstrated that an increase in waste loading of more than 20% could be achieved for this waste composition, and by extension higher loadings for wastes in the same group. An extended duration CCIM melter test was conducted on an AZ-101 waste simulant using the CCIM platform at the Idaho National Laboratory (INL). The melter was continually operated for approximately 80 hours demonstrating that the AZ-101 high waste loading glass composition could be readily processed using the CCIM technology. The resulting glass was close to the targeted composition and exhibited excellent durability in both

Transuranic solid waste that has been generated as a result of the production of nuclear material for the United States defense program at the Savannah River Site (SRS) has been stored in more than 30,000 55-gallon drums and various size carbon steel boxes since 1953. Nearly two thirds of those containers have been processed and shipped to the Waste Isolation Pilot Plant. Among the containers assayed so far, the results indicate several drums with fissile inventories significantly higher (600-1000 grams {sup 239}Pu) than their original assigned values. While part of this discrepancy can be attributed to the past limited assay capabilities, human errors are believed to be the primary contributor. This paper summarizes an assessment of the probability of occurrence of a criticality accident during handling of the remaining transuranic waste containers at SRS.

HighLevelWaste Programs has radioactive waste storage, treatment and processing facilities that are located in the F and H Areas at the Savannah River Site. These facilities include the Effluent Treatment Facility (ETF), F and H Area Tank Farms, Extended Sludge Processing (ESP), and In-Tank Precipitation (ITP). Job wastes are generated from operation, maintenance, and construction activities inside radiological areas. These items may have been contaminated with radioactive supernate, salt, and sludge material. Most of these wastes will be disposed of in the E-area Vaults. Therefore, an isotopic and hazardous characterization must be performed. The characterization of HLW supernate radionuclides is discussed in Chapter I. The characterization for salt and sludge phases, which can also contaminate LLW, will be included in other Chapters.

The potential for realizing cost savings in the disposal of defense high-levelwaste through process and design modificatins has been considered. Proposed modifications range from simple changes in the canister design to development of an advanced melter capable of processing glass with a higher waste loading. Preliminary calculations estimate the total disposal cost (not including capital or operating costs) for defense high-levelwaste to be about $7.9 billion dollars for the reference conditions described in this paper, while projected savings resulting from the proposed process and design changes could reduce the disposal cost of defense high-levelwaste by up to $5.2 billion.

This research investigated four techniques that could be applicable for mapping of solids remaining in radioactive waste tanks at the Savannah River Site: stereo vision, LIDAR, flash LIDAR, and Structure from Motion (SfM). Stereo vision is the least appropriate technique for the solids mapping application. Although the equipment cost is low and repackaging would be fairly simple, the algorithms to create a 3D image from stereo vision would require significant further development and may not even be applicable since stereo vision works by finding disparity in feature point locations from the images taken by the cameras. When minimal variation in visual texture exists for an area of interest, it becomes difficult for the software to detect correspondences for that object. SfM appears to be appropriate for solids mapping in waste tanks. However, equipment development would be required for positioning and movement of the camera in the tank space to enable capturing a sequence of images of the scene. Since SfM requires the identification of distinctive features and associates those features to their corresponding instantiations in the other image frames, mockup testing would be required to determine the applicability of SfM technology for mapping of waste in tanks. There may be too few features to track between image frame sequences to employ the SfM technology since uniform appearance may exist when viewing the remaining solids in the interior of the waste tanks. Although scanning LIDAR appears to be an adequate solution, the expense of the equipment ($80,000-$120,000) and the need for further development to allow tank deployment may prohibit utilizing this technology. The development would include repackaging of equipment to permit deployment through the 4-inch access ports and to keep the equipment relatively uncontaminated to allow use in additional tanks. 3D flash LIDAR has a number of advantages over stereo vision, scanning LIDAR, and SfM, including full frame

This report presents the results of the Independent Project Evaluation (IPE) Team assessment of the Westinghouse Savannah River Company High-LevelWaste Salt Disposition Systems Engineering (SE) Team's deliberations, evaluations, and selections. The Westinghouse Savannah River Company concluded in early 1998 that production goals and safety requirements for processing SRS HLW salt to remove Cs-137 could not be met in the existing In-Tank Precipitation Facility as currently configured for precipitation of cesium tetraphenylborate. The SE Team was chartered to evaluate and recommend an alternative(s) for processing the existing HLW salt to remove Cs-137. To replace the In-Tank Precipitation process, the Savannah River Site HLW Salt Disposition SE Team downselected (October 1998) 140 candidate separation technologies to two alternatives: Small-Tank Tetraphenylborate (TPB) Precipitation (primary alternative) and Crystalline Silicotitanate (CST) Nonelutable Ion Exchange (backup alternative). The IPE Team, commissioned by the Department of Energy, concurs that both alternatives are technically feasible and should meet all salt disposition requirements. But the IPE Team judges that the SE Team's qualitative criteria and judgments used in their downselection to a primary and a backup alternative do not clearly discriminate between the two alternatives. To properly choose between Small-Tank TPB and CST Ion Exchange for the primary alternative, the IPE Team suggests the following path forward: Complete all essential R and D activities for both alternatives and formulate an appropriate set of quantitative decision criteria that will be rigorously applied at the end of the R and D activities. Concurrent conceptual design activities should be limited to common elements of the alternatives.

This report presents a study of alternative system architectures to provide onsite interim storage for the immobilized high-levelwaste produced by the Tank Waste Remediation System (TWRS) privatization vendor. It examines the contract and program changes that have occurred and evaluates their impacts on the baseline immobilized high-levelwaste (IHLW) interim storage strategy. In addition, this report documents the recommended initial interim storage architecture and implementation path forward.

In the residual liquid solidification effort, the primary alternative studied is the wiped film evaporator approach to solidifying salt well pumped liquids and returning the molten material to single shell tanks for microwave final stabilization to a hard dry product. Both systems analysis and experimental work are proceeding to evaluate this approach. The primary alternative for in situ stabilization of in-tank wastes is microwave drying of wet salt cake and unpumped sludges. Experimental work was successfully conducted on a 1/12 scale tank containing wet synthetic salt cake. Related systems analysis of a full scale system was initiated.

The evaluation studies of the proposed repository for long-term storage of spent nuclear fuel and high-level nuclear waste at Yucca Mountain, Nevada, are underway. Fulfillment of the requirements for limiting dose to the public, which includes containment of the radioactive waste emplaced in the proposed repository and subsequent slow release of radionuclides from the Engineered Barrier System (EBS) into the geosphere, will rely on a robust waste container design, among other EBS components. Part of the evaluation process involves sensitivity studies aimed at elucidating which model parameters contribute most to the waste package and overlying drip shield degradation characteristics. The model parameters identified for this study include (1) general corrosion rate parameters and (2) stress corrosion cracking (SCC) parameters. Temperature dependence and parameter uncertainty are evaluated for the general corrosion rate model parameters while for the SCC model parameters, uncertainty treatment of stress intensity factor, crack initiation threshold, and manufacturing flaw orientations are evaluated. Based on these evaluations new uncertainty distributions are generated and recommended for future analyses. Also, early waste package failures due to improper heat treatment were added to the waste package degradation model. The results of these investigations indicate that the waste package failure profiles are governed by the manufacturing flaw orientation model parameters. PMID:15078313

This report deals with technical information that is considered essential for demonstrating the ability of the high-level radioactive waste package to provide substantially complete containment'' of its contents (vitrified waste form or spent light-water reactor fuel) for a period of 300 to 1000 years in a geological repository environment. The discussion is centered around technical considerations of the repository environment, materials and fabrication processes for the waste package components, various degradation modes of the materials of construction of the waste packages, and inspection and monitoring of the waste package during the preclosure and retrievability period, which could begin up to 50 years after initiation of waste emplacement. The emphasis in this report is on metallic materials. However, brief references have been made to other materials such as ceramics, graphite, bonded ceramic-metal systems, and other types of composites. The content of this report was presented to an external peer review panel of nine members at a workshop held at the Center for Nuclear Waste Regulatory Analyses (CNWRA), Southwest Research Institute, San Antonio, Texas, April 2--4, 1990. The recommendations of the peer review panel have been incorporated in this report. There are two companion reports; the second report in the series provides state-of-the-art techniques for uncertainty evaluations. 97 refs., 1 fig.

The purpose of this calculation is to provide a bounding dose consequence analysis of the immobilized plutonium (can-in-canister) waste form to be handled at the Monitored Geologic Repository (MGR) at Yucca Mountain. The current concept for the Plutonium Can-in-Canister waste form is provided in Attachment III. A typical design basis event (DBE) defines a scenario that generally includes an initiating event and the sequences of events that follow. This analysis will provide (1) radiological releases and dose consequences for a postulated, bounding DBE and (2) design-related assumptions on which the calculated dose consequences are based. This analysis is part of the safety design basis for the repository. Results will be used in other analyses to determine or modify the safety classification and quality assurance level of repository structures, systems, and components (SSCs). The Quality Assurance (QA) program applies to this calculation. The work reported in this document is part of the analysis of MGR DBEs and is performed using AP-3.12Q, Calculations. The work done for this analysis was evaluated according to QAP-2-0, Control of Activities. This evaluation determined that such activities are subject to DOE/RW/0333PY Quality Assurance Requirements and Description (DOE 1998), requirements. This calculation is quality affecting because the results may be used to support analyses of repository SSCs per QAP-2-3, Classification of Permanent Items.

Glass and glass-ceramic forms developed for the immobilization of calcined high-levelwastes generated by Idaho Chemical Processing Plant (ICPP) fuel reprocessing activities have been investigated for ability to immobilize plutonium and to simultaneously incorporate calcined waste as an anti-proliferation barrier. Within the forms investigated, crystallization of host phases result in an increased loading of plutonium as well as its incorporation into potentially more durable phases than the glass. The host phases were initially formed and characterized with cerium (Ce{sup +4}) as a surrogate for plutonium (Pu{sup +4}) and samarium as a neutron absorber for criticality control. Verification of the surrogate testing results were then performed replacing cerium with plutonium. All testing was performed with surrogate calcined high-levelwaste. The results of these tests indicated that a potentially useful host phase, based on zirconia, can be formed either by devitrification or solid state reaction in the glass studied. This phase incorporates plutonium as well as samarium and the calcined waste becomes part of the matrix. Its ease of formation makes it potentially useful in excess plutonium dispositioning. Other durable host phases for plutonium and samarium, including zirconolite and zircon have been formed from zirconia or alumina calcine through cold press-sintering techniques and hot isostatic pressing. Host phase formation experiments conducted through vitrification or by cold press-sintering techniques are described and the results discussed. Recommendations are given for future work that extends the results of this study.

High-levelwaste (HLW) throughput (i.e., the amount of waste processed per unit time) is a function of two critical parameters: waste loading (WL) and melt rate. For the Waste Treatment and Immobilization Plant (WTP) at the Hanford Site and the Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS), increasing HLW throughput would significantly reduce the overall mission life cycle costs for the Department of Energy (DOE). It has been proposed that a team of glass formulation and processing experts at the Pacific Northwest National Laboratory (PNNL), Savannah River National Laboratory (SRNL), and Vitreous State Laboratory (VSL) at Catholic University of America develop a systematic approach to increase HLW throughput (by increasing WL with minimal or positive impacts on melt rate). Programmatically, this task is aimed at proof-of-principle testing and the development of tools to improve waste loading and melt rate, which will lead to higher waste throughput. The following four specific tasks have been proposed to meet this programmatic objective: (1) Integration and Oversight, (2) Crystal Accumulation Modeling (led by PNNL)/Higher Waste Loading Glasses (led by SRNL), (3) Melt Rate Evaluation and Modeling, and (4) Melter Scale Demonstrations. The details of these tasks can be found in the associated task plan WSRC-STI-2007-00483. The current study is focused on Task 2 (crystal accumulation modeling and higher waste loading glasses) and involves glass formulation and physical property testing by both PNNL and SRNL (as defined in the PNNL and SRNL test plans). The intent of this report is to document the chemical composition and Product Consistency Test (PCT) results and statistical analysis of PNNL's Test Matrix 1 glasses. Note that this document is only a compilation of the data collected by SRNL for PNNL's glasses in support of this task and no conclusions will be drawn.

Pretreatment processes at the Savannah River Site will separate {sup 90}Sr, alpha-emitting and radionuclides (i.e., actinides) and {sup 137}Cs prior to disposal of the high-level nuclear waste. Separation of {sup 90}Sr and alpha-emitting radionuclides occurs by ion exchange/adsorption using an inorganic material, monosodium titanate (MST). Previously reported testing with simulants indicates that the MST exhibits high selectivity for strontium and actinides in high ionic strength and strongly alkaline salt solutions. This paper provides a summary of data acquired to measure the performance of MST to remove strontium and actinides from actual waste solutions. These tests evaluated the effects of ionic strength, mixing, elevated alpha activities, and multiple contacts of the waste with MST. Tests also provided confirmation that MST performs well at much larger laboratory scales (300-700 times larger) and exhibits little affinity for desorption of strontium and plutonium during washing.

The solubility of plutonium in a Sludge Batch 6 (SB6) reference glass and the effect of incorporation of Pu in the glass on specific glass properties were evaluated. A Pu loading of 1 wt % in glass was studied. Prior to actual plutonium glass testing, surrogate testing (using Hf as a surrogate for Pu) was conducted to evaluate the homogeneity of significant quantities of Hf (Pu) in the glass, determine the most appropriate methods to evaluate homogeneity for Pu glass testing, and to evaluate the impact of Hf loading in the glass on select glass properties. Surrogate testing was conducted using Hf to represent between 0 and 1 wt % Pu in glass on an equivalent molar basis. A Pu loading of 1 wt % in glass translated to {approx}18 kg Pu per Defense Waste Processing Facility (DWPF) canister, or about 10X the current allowed limit per the Waste Acceptance Product Specifications (2500 g/m{sup 3} of glass or about 1700 g/canister) and about 30X the current allowable concentration based on the fissile material concentration limit referenced in the Yucca Mountain Project License Application (897 g/m{sup 3}3 of glass or about 600 g Pu/canister). Based on historical process throughput data, this level was considered to represent a reasonable upper bound for Pu loading based on the ability to provide Pu containing feed to the DWPF. The task elements included evaluating the distribution of Pu in the glass (e.g. homogeneity), evaluating crystallization within the glass, evaluating select glass properties (with surrogates), and evaluating durability using the Product Consistency Test -- Method A (PCT-A). The behavior of Pu in the melter was evaluated using paper studies and corresponding analyses of DWPF melter pour samples.The results of the testing indicated that at 1 wt % Pu in the glass, the Pu was homogeneously distributed and did not result in any formation of plutonium-containing crystalline phases as long as the glass was prepared under 'well-mixed' conditions. The

A series of static leach tests was conducted using glasses developed for vitrifying tank wastes at the Savannah River Site to monitor the disposition of actinide elements upon corrosion of the glasses. In these tests, glasses produced from SRL 131 and SRL 202 frits were corroded at 90{degrees}C in a tuff groundwater. Tests were conducted using crushed glass at different glass surface area-to-solution volume (S/V) ratios to assess the effect of the S/V on the solution chemistry, the corrosion of the glass, and the disposition of actinide elements. Observations regarding the effects of the S/V on the solution chemistry and the corrosion of the glass matrix have been reported previously. This paper highlights the solution analyses performed to assess how the S/V used in a static leach test affects the disposition of actinide elements between fractions that are suspended or dissolved in the solution, and retained by the altered glass or other materials.

Los Alamos National Laboratory has performed a comprehensive probabilistic safety assessment (PSA) that includes consideration of external events for the weapons-production wastes stored in tank number 241-SY-101, commonly known as Tank 101-SY, as configured in December 1992. This tank, which periodically releases (``burps``) a gaseous mixture of hydrogen, nitrous oxide, ammonia, and nitrogen, was analyzed because of public safety concerns associated with the potential for release of radioactive tank contents should this gas mixture be ignited during one of the burps. In an effort to mitigate the burping phenomenon, an experiment is underway in which a large pump has been inserted into the tank to determine if pump-induced circulation of the tank contents will promote a slow, controlled release of the gases. This PSA for Tank 101-SY, which did not consider the pump experiment or future tank-remediation activities, involved three distinct tasks. First, the accident sequence analysis identified and quantified those potential accidents whose consequences result in tank material release. Second, characteristics and release paths for the airborne and liquid radioactive source terms were determined. Finally, the consequences, primarily onsite and offsite potential health effects resulting from radionuclide release, were estimated, and overall risk curves were constructed. An overview of each of these tasks and a summary of the overall results of the analysis are presented in the following sections.

The overall objective of this program, SRD-99-08, was to design and test suitable materials, which can be used to close high-levelwaste tanks at SRS. Fill materials can be designed to perform several functions including chemical stabilization and/or physical encapsulation of incidental waste so that the potential for transport of contaminants into the environment is reduced. Also they are needed to physically stabilize the void volume in the tanks to prevent/minimize future subsidence and inadvertent intrusion. The intent of this work was to develop a zero-bleed soil CLSM (ZBS-CLSM) and a zero-bleed concrete mix (ZBC) which meet the unique placement and stabilization/encapsulation requirements for high-levelwaste tank closures. These mixes in addition to the zero-bleed CLSM mixes formulated for closure of Tanks 17-F and 20-F provide design engineers with a suite of options for specifying materials for future tank closures.

Nickel ferrocyanide compounds (Na{sub 2-x}Cs{sub x}NiFe (CN){sub 6}) were produced in a scavenging process to remove {sup 137}Cs from Hanford Site single-shell tank waste supernates. Methods for determining total cyanide in Hanford Site high-levelwastes are needed for the evaluation of potential exothermic reactions between cyanide and oxidizers such as nitrate and for safe storage, processing, and management of the wastes in compliance with regulatory requirements. Hanford Site laboratory experience in determining cyanide in high-levelwastes is summarized. Modifications were made to standard cyanide methods to permit improved handling of high-levelwaste samples and to eliminate interferences found in Hanford Site waste matrices. Interferences and associated procedure modifications caused by high nitrates/nitrite concentrations, insoluble nickel ferrocyanides, and organic complexants are described.

The failure frequency of Type I and Type II HighLevelWaste tanks was calculated. The degradation mechanism that could lead to large break failure and the credits taken for steps taken to prevent large break failure were considered.

This report details the experimental effort to demonstrate the continuous precipitation of cesium from Savannah River Site HighLevelWaste using sodium tetraphenylborate. In addition, the experiments examined the removal of strontium and various actinides through addition of monosodium titanate.

The HLW glass was produced from a HLW sludge slurry (Envelope D Waste), eluate waste streams containing highlevels of Cs-137 and Tc-99, solids containing both Sr-90 and transuranics (TRU), and glass-forming chemicals. The eluates and Sr-90/TRU solids were obtained from ion-exchange and precipitation pretreatments, respectively, of other Hanford supernate samples (Envelopes A, B and C Waste). The glass was vitrified by mixing the different waste streams with glass-forming chemicals in platinum/gold crucibles and heating the mixture to 1150 degree C. Resulting glass analyses indicated that the HLW glass waste form composition was close to the target composition. The targeted waste loading of Envelope D sludge solids in the HLW glass was 30.7 wt percent, exclusive of Na and Si oxides. Condensate samples from the off-gas condenser and off-gas dry-ice trap indicated that very little of the radionuclides were volatilized during vitrification. Microstructure analysis of the HLW glass using Scanning Electron Microscopy (SEM) and Energy Dispersive X-Ray Analysis (EDAX) showed what appeared to be iron spinel in the HLW glass. Further X-Ray Diffraction (XRD) analysis confirmed the presence of nickel spinel trevorite (NiFe2O4). These crystals did not degrade the leaching characteristics of the glass. The HLW glass waste form passed leach tests that included a standard 90 degree C Product Consistency Test (PCT) and a modified version of the United States Environmental Protection Agency Toxicity Characteristic Leaching Procedure (TCLP).

Designs of a three-column carousel process based on crystalline silicotitanate (CST) ion exchanger have been developed for removing radioactive 137Cs+ from Savannah River Site's (SRS) nuclear wastes. A multicomponent ion exchange equilibrium model (Zheng et al., 1997) from Texas A&M University, which is based on batch data obtained from CST powder, is used to generate cesium loading data at different cesium concentrations for various types of SRSwastes. These loading data are fit to the Langmuir equation to obtain effective single-component cesium isotherm parameters. The predictions are in reasonable agreement with batch test data obtained from CST powder, an early CST pellet batch (38B), and a later batch (IE911) using two SRSwaste simulants. The ratios between experimental cesium distribution coefficients and predicted values are between 0.56 and 1.0. The variation appears to be due to inadequate equilibration time in some of the batches. Mass transfer parameters are estimated by analyzing column data of a simulated SRSwaste and Melton Valley Storage Tank W29 (MVST-W29) waste. The intraparticle diffusivity estimated for the two wastes can be well correlated by means of the Stokes-Einstein equation.Simulations are performed to determine the length of the mass transfer zone for given feed compositions, Cs+ concentrations, and linear velocities. In order to ensure high column utilization during both the transient and cyclic steady state periods, the length of a single segment in the carousel process is chosen to be the mass transfer zone length after the concentration wave achieves a constant pattern. Analysis of the dimensionless groups in the differential mass balance equations reveals that the normalized mass transfer zone length is linearly proportional to the particle Peclet number. The proportionality constant is a function of the waste composition and the Cs+ concentration in the waste. The higher the effective Cs+ capacity and the higher the Cs

The report contains two major sections: Section I - An Improved Polyphase Ceramic for High-Level Defense Nucleation Waste reports the work conducted on titanium-silica based ceramics for immobilizing Savannah River Plant waste. Section II - Formulation and Processing of Alumina Based Ceramic Nuclear Waste Forms describes the work conducted on developing a generic alumina and alumina-silica based ceramic waste form capable of immobilizing any nuclear waste with a high aluminum content. Such wastes include the Savannah River Plant wastes, Hanford neutralized purex wastes, and Hanford N-Reactor acid wastes. The design approach and process technology in the two reports demonstrate how the generic high waste loaded ceramic form can be applied to a broad range of nuclear waste compositions. The individual sections are abstracted and indexed separately.

The Deflagration Pressure Analysis Code (DPAC) was developed primarily to evaluate peak pressures for deflagrations in radioactive waste storage and process facilities at the Savannah River Site (SRS). Deflagrations in these facilities are generally considered to be incredible events, but it was judged prudent to develop modeling capabilities in order to facilitate risk estimates. DPAC is essentially an engineering analysis tool, as opposed to a detailed thermal hydraulics code. It accounts for mass loss via venting, energy dissipation by radiative heat transfer, and gas PdV work. Volume increases due to vessel deformation can also be included using pressure-volume data from a structural analysis of the enclosure. This paper presents an overview of the code, benchmarking, and applications at SRS.

This paper presents the results of glass formulation development and melter testing to identify high waste loading glasses to treat high-Al highlevelwaste (HLW) at Hanford. Previous glass formulations developed for this HLW had high waste loadings but their processing rates were lower that desired. The present work was aimed at improving the glass processing rate while maintaining high waste loadings. Glass formulations were designed, prepared at crucible-scale and characterized to determine their properties relevant to processing and product quality. Glass formulations that met these requirements were screened for melt rates using small-scale tests. The small-scale melt rate screening included vertical gradient furnace (VGF) and direct feed consumption (DFC) melter tests. Based on the results of these tests, modified glass formulations were developed and selected for larger scale melter tests to determine their processing rate. Melter tests were conducted on the DuraMelter 100 (DMIOO) with a melt surface area of 0.11 m{sup 2} and the DuraMelter 1200 (DMI200) HLW Pilot Melter with a melt surface area of 1.2 m{sup 2}. The newly developed glass formulations had waste loadings as high as 50 wt%, with corresponding Al{sub 2}O{sub 3} concentration in the glass of 26.63 wt%. The new glass formulations showed glass production rates as high as 1900 kg/(m{sup 2}.day) under nominal melter operating conditions. The demonstrated glass production rates are much higher than the current requirement of 800 kg/(m{sup 2}.day) and anticipated future enhanced Hanford Tank Waste Treatment and Immobilization Plant (WTP) requirement of 1000 kg/(m{sup 2}.day).

In July of 1986, the first in-situ test involving burial of simulated high-levelwaste (HLW) forms conducted in the United States was started. This program, called the Materials Interface Interactions Test or MIIT, comprises the largest, most cooperative field-testing venture in the international waste management community. In July of 1991, the experimental portion of the 5-year MIIT study was completed on schedule. During this time interval, many in-situ measurements were performed, thousands of brine analyses conducted, and hundreds of waste glass and package components exhumed and evaluated after 6 mo., 1 yr., 2 yr. and 5 yr. burial periods. Although analyses are still in progress, the performance of SRSwaste glass based on all data currently available has been seen to be excellent thus far. Initial analyses and assessment of Savannah River (SR) waste glass after burial in WIPP at 90{degrees}C for 5 years are presented in this document.

In July of 1986, the first in-situ test involving burial of simulated high-levelwaste (HLW) forms conducted in the United States was started. This program, called the Materials Interface Interactions Test or MIIT, comprises the largest, most cooperative field-testing venture in the international waste management community. In July of 1991, the experimental portion of the 5-year MIIT study was completed on schedule. During this time interval, many in-situ measurements were performed, thousands of brine analyses conducted, and hundreds of waste glass and package components exhumed and evaluated after 6 mo., 1 yr., 2 yr. and 5 yr. burial periods. Although analyses are still in progress, the performance of SRSwaste glass based on all data currently available has been seen to be excellent thus far. Initial analyses and assessment of Savannah River (SR) waste glass after burial in WIPP at 90{degrees}C for 5 years are presented in this document.

continue to accomplish DOE’s critical nuclear material missions (e.g., processing in H-Canyon and plutonium storage in K-Area). These demonstrations can be accomplished in a more cost-effective manner through the use of existing facilities in conjunction with ongoing missions. Essentially, the R&D program would not need to pay the full operational cost of a facility, just the incremental cost of performing the demonstration. Current Center activities have been focused on integrating advanced safeguards monitoring technology demonstrations into the SRS H-Canyon and advanced location technology demonstrations into K-Area Materials Storage. These demonstrations are providing valuable information to researchers and program owners. In addition these demonstrations are providing the Center with an improved protocol for demonstration management that can be exercised across the entire SRS (and to offsite venues) to ensure that future demonstrations are done efficiently and provide an opportunity to use these unique assets for multiple purposes involving national laboratories, academia, and commercial entities. Key among the envisioned future use of SRS assets is the demonstration of new radioactive waste management technologies critical for advancing the mission needs of the DOE-EM program offices in their efforts to cleanup 107 sites across the United States. Of particular interest is the demonstration of separations technologies in H-Canyon. Given the modular design of H-Canyon, those demonstrations would be accomplished using a process frame. The demonstration equipment would be installed on the process frame and that frame would then be positioned into an H-Canyon cell so that the demonstration is performed in a radiological environment involving prototypic nuclear materials.

Discusses some of the problems associated with the safe disposal of high-level nuclear wastes. Describes several waste disposal plans developed by various nations. Outlines the multiple-barrier concept of isolation in deep geological questions associated with the implementation of such a method. (TW)

The objective of this work is to develop a more detailed mechanistic understanding of the thermal reactions that lead to gas production in certain high-levelwaste storage tanks at the Hanford, Washington site. Prediction of the combustion hazard for these wastes and engineering ...

Efforts are underway to qualify the Next-Generation Solvent for the Caustic Side Solvent Extraction (CSSX) process. Researchers at multiple national laboratories have been involved in this effort. As part of the effort to qualify the solvent extraction system at the Savannah River Site (SRS), SRNL performed a number of tests at various scales. First, SRNL completed a series of batch equilibrium, or Extraction-Scrub-Strip (ESS), tests. These tests used {approx}30 mL of Next-Generation Solvent and either actual SRS tank waste, or waste simulant solutions. The results from these cesium mass transfer tests were used to predict solvent behavior under a number of conditions. At a larger scale, SRNL assembled 12 stages of 2-cm (diameter) centrifugal contactors. This rack of contactors is structurally similar to one tested in 2001 during the demonstration of the baseline CSSX process. Assembly and mechanical testing found no issues. SRNL performed a nonradiological test using 35 L of cesium-spiked caustic waste simulant and 39 L of actual tank waste. Test results are discussed; particularly those related to the effectiveness of extraction.

The calculation method of neutron yield in the (α, n) reaction for a homogeneous material of arbitrary composition is represented. It is shown that the use of the ORIGEN 2 code excluding the real elemental composition of vitrified high-levelwaste leads to significant underestimation of the neutron yield in the (α, n) reaction. For vitrified high-levelwaste and spent nuclear fuel from VVER, the neutron fluxes are analyzed. The thickness of the protective materials for a transfer cask and a shipping cask with vitrified highlevel waste are estimated.

The calculation method of neutron yield in the (α, n) reaction for a homogeneous material of arbitrary composition is represented. It is shown that the use of the ORIGEN 2 code excluding the real elemental composition of vitrified high-levelwaste leads to significant underestimation of the neutron yield in the (α, n) reaction. For vitrified high-levelwaste and spent nuclear fuel from VVER, the neutron fluxes are analyzed. The thickness of the protective materials for a transfer cask and a shipping cask with vitrified highlevel waste are estimated.

In 1980, the US Department of Energy (DOE) established the West Valley Solidification Project as the result of legislation passed by the US Congress. The purpose of this project was to carry out a highlevel nuclear waste management demonstration project at the Western New York Nuclear Service Center in West Valley, New York. The DOE authorized the Pacific Northwest Laboratory (PNL), which is operated by Battelle Memorial Institute, to assess alternative processes for treatment and solidification of the WNYNSC high-levelwastes. The Process Alternatives Study is the suject of this report. Two pretreatment approaches and several waste form processes were selected for evaluation in this study. The two waste treatment approaches were the salt/sludge separation process and the combined waste process. Both terminal and interim waste form processes were studied.

The Department of Energy (DOE) has prepared an environmental assessment (EA) (DOEEA-1164) for the proposed closure of the high-levelwaste tanks in F- and H-Areas on the Savannah River Site (SRS), near Aiken, South Carolina. Based on the analyses in the EA, DOE has determined that the proposed action is not a major Federal action significantly affecting the quality of the human environment within the meaning of the National Environmental Policy Act (NEPA) of 1969. Therefore, the preparation of an environmental impact statement is not required, and DOE is issuing this Finding of No Significant Impact (FONSI).

This document summarizes work performed at participating sites on the immobilization of high-levelwastes from the chemical reprocessing of reactor fuels. The plan is to develop waste form alternatives for each of the three DOE sites (SRP, ICPP, and Hanford). Progress is reported in the following areas: waste preparation; fixation in glass, concrete, tailored ceramics, and coated particles; process and equipment development; and final handling. 12 figures, 19 tables. (DLC)

High-level nuclear waste is being vitrified, i.e., converted to a durable glass that can be stored in a safe repository for hundreds of thousands of years. Waste vitrification is accomplished in reactors called melters to which the waste is charged together with glass-forming additives. The mixture is electrically heated to a temperature as high as 1150 decrees C to create a melt that becomes glass on cooling.

This criticality safety evaluation covers operations for waste in underground storage tanks at the high-levelwaste tank farms on the Hanford site. This evaluation provides the bases for criticality safety limits and controls to govern receipt, transfer, and long-term storage of tank waste. Justification is provided that a nuclear criticality accident cannot occur for tank farms operations, based on current fissile material and operating conditions.

This paper provides an update from last year and describes project successes and issues associated with the management and work required to safely store, enhance readiness for waste feed delivery, and prepare for treated waste receipts for the approximately 53 million gallons of mixed and high-levelwaste currently in aging tanks at the Hanford Site. The Hanford Site is a 560 square-mile area in southeastern Washington State near Richland, Washington.

High-level radioactive wastes generated from nuclear fuel reprocessing require treatment for effective long-term storage. Heating by microwave energy is explored in processing of two possible waste forms: (1) drying of a pelleted form of calcined waste; and (2) vitrification of calcined waste. It is shown that residence times for these processes can be greatly reduced when using microwave energy rather than conventional heating sources, without affecting product properties. Compounds in the waste and in the glass frit additives couple very well with the 2.45 GHz microwave field so that no special microwave absorbers are necessary.

A new approach, termed SIPS (Small In-Tank Processing System), that enables the in-tank processing and separation of high-level tank wastes into high-levelwaste (HLW) and low-level waste (LLW) streams that are suitable for vitrification, is described. Presently proposed pretreatment systems, such as enhanced sludge washing (ESW) and TRUEX, require that the high-level tank wastes be retrieved and pumped to a large, centralized processing facility, where the various waste components are separated into a relatively small, radioactively concentrated stream (HLW), and a relatively large, predominantly non-radioactive stream (LLW). In SIPS, a small process module, typically on the order of 1 meter in diameter and 4 meters in length, is inserted into a tank. During a period of approximately six months, it processes the solid/liquid materials in the tank, separating them into liquid HLW and liquid LLW output streams that are pumped away in two small diameter (typically 3 cm o.d.) pipes. The SIPS concept appears attractive for pretreating highlevelwastes, since it would: (1) process waste in-situ in the tanks, (2) be cheaper and more reliable than a larger centralized facility, (3) be quickly demonstrable at full scale, (4) have less technical risk, (5) avoid having to transfer unstable slurries for long distances, and (6) be simple to decommission and dispose of. Further investigation of the SIPS concept appears desirable, including experimental testing and development of subscale demonstration units.

This report presents a proposal for quantitative and generally applicable risk-based definitions of high-level and other radioactive wastes. On the basis of historical descriptions and definitions of high-levelwaste (HLW), in which HLW has been defined in terms of its source as waste from reprocessing of spent nuclear fuel, we propose a more general definition based on the concept that HLW has two distinct attributes: HLW is (1) highly radioactive and (2) requires permanent isolation. This concept leads to a two-dimensional waste classification system in which one axis, related to ''requires permanent isolation,'' is associated with long-term risks from waste disposal and the other axis, related to ''highly radioactive,'' is associated with shorter-term risks due to highlevels of decay heat and external radiation. We define wastes that require permanent isolation as wastes with concentrations of radionuclides exceeding the Class-C limits that are generally acceptable for near-surface land disposal, as specified in the US Nuclear Regulatory Commission's rulemaking 10 CFR Part 61 and its supporting documentation. HLW then is waste requiring permanent isolation that also is highly radioactive, and we define ''highly radioactive'' as a decay heat (power density) in the waste greater than 50 W/m/sup 3/ or an external radiation dose rate at a distance of 1 m from the waste greater than 100 rem/h (1 Sv/h), whichever is the more restrictive. This proposal also results in a definition of Transuranic (TRU) Waste and Equivalent as waste that requires permanent isolation but is not highly radioactive and a definition of low-level waste (LLW) as waste that does not require permanent isolation without regard to whether or not it is highly radioactive.

Performance modeling for Low Level Mixed Waste disposal was conducted using the measured leach rates from a number of vitrified waste formulations. The objective of the study was to determine if the improved durability of a vitrified mixed waste would allow trench disposal at the Savannah River Site (SRS). Leaching data were compiled from twenty-nine diverse reference glasses, encompassing a wide range of exposed glass surface area to leachant volume ratios (SA/V), and various leachant solutions; all of which had been leached at 90 degrees Celsius, using the MCC-1 or PCT procedures (ASTM Procedures C1220-92 and C1285-94, respectively). The normalized leach rates were scaled to the ambient disposal temperature of 25 degrees Celsius, and compared to the allowable leach rate of uranium - which would meet the performance assessment requirements. The results indicated that a glass of above average durability (vs. the reference glasses) would meet the uranium leaching concentration for direct SRS trench disposal.

High-level nuclear waste is being immobilized at the Savannah River Site (SRS) by vitrification into borosilicate glass at the Defense Waste Processing Facility (DWPF). Control of the Reduction/Oxidation (REDOX) equilibrium in the DWPF melter is critical for processing highlevel liquid wastes. Foaming, cold cap roll-overs, and off-gas surges all have an impact on pouring and melt rate during processing of high-levelwaste (HLW) glass. All of these phenomena can impact waste throughput and attainment in Joule heated melters such as the DWPF. These phenomena are caused by gas-glass disequilibrium when components in the melter feeds convert to glass and liberate gases such as H{sub 2}O vapor (steam), CO{sub 2}, O{sub 2}, H{sub 2}, NO{sub x}, and/or N{sub 2}. During the feed-to-glass conversion in the DWPF melter, multiple types of reactions occur in the cold cap and in the melt pool that release gaseous products. The various gaseous products can cause foaming at the melt pool surface. Foaming should be avoided as much as possible because an insulative layer of foam on the melt surface retards heat transfer to the cold cap and results in low melt rates. Uncontrolled foaming can also result in a blockage of critical melter or melter off-gas components. Foaming can also increase the potential for melter pressure surges, which would then make it difficult to maintain a constant pressure differential between the DWPF melter and the pour spout. Pressure surges can cause erratic pour streams and possible pluggage of the bellows as well. For these reasons, the DWPF uses a REDOX strategy and controls the melt REDOX between 0.09 {le} Fe{sup 2+}/{summation}Fe {le} 0.33. Controlling the DWPF melter at an equilibrium of Fe{sup +2}/{summation}Fe {le} 0.33 prevents metallic and sulfide rich species from forming nodules that can accumulate on the floor of the melter. Control of foaming, due to deoxygenation of manganic species, is achieved by converting oxidized MnO{sub 2} or Mn

Pyrochemical methods are being evaluated for the separation of actinides and fission products from inert material in high-levelwaste calcine. Separation processes have the potential of reducing waste disposal costs by reducing the volume of high-levelwaste requiring final disposal in a repository. Tests were conducted to evaluate high temperature volatilization for removing {sup 137}Cs from four types of calcines. The results for pilot plant calcines indicate greater than 99% cesium removal for alumina and fluorinel/sodium calcines heated at 1000{degree}C and 99% removal for zirconia calcine heated at 1100{degree}C. Tests with actual calcine generated at the New Waste Calcining Facility (NWCF) from a blend of aluminium, sodium and zirconium/fluorinel wastes resulted in 96% cesium removal at 1000{degree}C and greater than 99% removal at 1170{degree}C.

The Replacement HighLevelWaste Evaporator Project was conceived in 1985 to reduce the volume of the highlevel radioactive waste Process of the highlevelwaste has been accomplished up to this time using Bent Tube type evaporators and therefore, that type evaporator was selected for this project. The Title I Design of the project was 70% completed in late 1990. The Department of Energy at that time hired an independent consulting firm to perform a complete review of the project. The DOE placed a STOP ORDER on purchasing the evaporator in January 1991. Essentially, no construction was to be done on this project until all findings and concerns dealing with the type and design of the evaporator are resolved. This report addresses two aspects of the DOE design review; (1) Comparing the Bent Tube Evaporator with the Forced Circulation Evaporator, (2) The design portion of the DOE Project Review - concentrated on the mechanical design properties of the evaporator. 1 ref.

The Replacement HighLevelWaste Evaporator Project was conceived in 1985 to reduce the volume of the highlevel radioactive waste Process of the highlevelwaste has been accomplished up to this time using Bent Tube type evaporators and therefore, that type evaporator was selected for this project. The Title I Design of the project was 70% completed in late 1990. The Department of Energy at that time hired an independent consulting firm to perform a complete review of the project. The DOE placed a STOP ORDER on purchasing the evaporator in January 1991. Essentially, no construction was to be done on this project until all findings and concerns dealing with the type and design of the evaporator are resolved. This report addresses two aspects of the DOE design review; (1) Comparing the Bent Tube Evaporator with the Forced Circulation Evaporator, (2) The design portion of the DOE Project Review - concentrated on the mechanical design properties of the evaporator. 1 ref.

Near-field temperatures resulting from the storage of high-levelwaste canisters and spent unreprocessed fuel assembly canisters in geologic formations were determined. Preliminary design of the repository was modeled for a heat transfer computer code, HEATING5, which used the finite difference method to evaluate transient heat transfer. The heat transfer system was evaluated with several two and three dimensional models which transfer heat by a combination of conduction, natural convention, and radiation. Physical properties of the materials in the model were based upon experimental values for the various geologic formations. The effects of canister spacing, fuel age, and use of an overpack were studied for the analysis of the spent fuel canisters; salt, granite, and basalt were considered as the storage media. The effects of canister diameter and use of an overpack were studied for the analysis of the high-levelwaste canisters; salt was considered as the only storage media for high-levelwaste canisters.

Removing phosphate from alkaline high-levelwaste sludges at the Department of Energy's Hanford Site in Washington State is necessary to increase the waste loading in the borosilicate glass waste form that will be used to immobilize the highly radioactive fraction of these wastes. We are developing a process which first leaches phosphate from the high-levelwaste solids with aqueous sodium hydroxide, and then isolates the phosphate by precipitation with calcium oxide. Tests with actual tank waste confirmed that this process is an effective method of phosphate removal from the sludge and offers an additional option for managing the phosphorus in the Hanford tank waste solids. The presence of vibrationally active species, such as nitrate and phosphate ions, in the tank waste processing streams makes the phosphate removal process an ideal candidate for monitoring by Raman or infrared spectroscopic means. As a proof-of-principle demonstration, Raman and Fourier transform infrared (FTIR) spectra were acquired for all phases during a test of the process with actual tank waste. Quantitative determination of phosphate, nitrate, and sulfate in the liquid phases was achieved by Raman spectroscopy, demonstrating the applicability of Raman spectroscopy for the monitoring of these species in the tank waste process streams.

Removing phosphate from alkaline high-levelwaste sludges at the Department of Energy's Hanford Site in Washington State is necessary to increase the waste loading in the borosilicate glass waste form that will be used to immobilize the highly radioactive fraction of these wastes. We are developing a process which first leaches phosphate from the high-levelwaste solids with aqueous sodium hydroxide, and then isolates the phosphate by precipitation with calcium oxide. Tests with actual tank waste confirmed that this process is an effective method of phosphate removal from the sludge and offers an additional option for managing the phosphorus in the Hanford tank waste solids. The presence of vibrationally active species, such as nitrate and phosphate ions, in the tank waste processing streams makes the phosphate removal process an ideal candidate for monitoring by Raman or infrared spectroscopic means. As a proof-of-principle demonstration, Raman and Fourier transform infrared (FTIR) spectra were acquired for all phases during a test of the process with actual tank waste. Quantitative determination of phosphate, nitrate, and sulfate in the liquid phases was achieved by Raman spectroscopy, demonstrating the applicability of Raman spectroscopy for the monitoring of these species in the tank waste process streams. PMID:22571620

Yucca Mountain in Nevada represents the proposed solution to what has been a lengthy national effort to dispose of high-level radioactive waste, waste which must be isolated from the biosphere for tens of thousands of years. This chapter reviews the background of that national effort and includes some discussion of international work in order to provide a more complete framework for the problem of waste disposal. Other chapters provide the regional geologic setting, the geology of the Yucca Mountain site, the tectonics, and climate (past, present, and future). These last two chapters are integral to prediction of long-term waste isolation.

Yucca Mountain in Nevada represents the proposed solution to what has been a lengthy national effort to dispose of high-level radioactive waste, waste which must be isolated from the biosphere for tens of thousands of years. This chapter reviews the background of that national effort and includes some discussion of international work in order to provide a more complete framework for the problem of waste disposal. Other chapters provide the regional geologic setting, the geology of the Yucca Mountain site, the tectonics, and climate (past, present, and future). These last two chapters are integral to prediction of long-term waste isolation. ?? 2007 Geological Society of America. All rights reserved.

The electrometallurgical treatment of metallic spent nuclear fuel produces two high-levelwaste streams; cladding hulls and chloride salt. Argonne National Laboratory is developing a glass bonded sodalite waste form to immobilize the salt waste stream. The waste form consists of 75 Vol.% crystalline sodalite (containing the salt) with 25 Vol.% of an ''intergranular'' glassy phase. Microindentation fracture toughness measurements were performed on representative samples of this material using a Vickers indenter. Palmqvist cracking was confirmed by post-indentation polishing of a test sample. Young's modulus was measured by an acoustic technique. Fracture toughness, microhardness, and Young's modulus values are reported, along with results from scanning electron microscopy studies.

Vitrification is currently the most widely used technology for the treatment of highlevel radioactive wastes (HLW) throughout the world. At the Savannah River Site (SRS) actual HLW tank waste has successfully been processed to stringent product and process constraints without any rework into a stable borosilicate glass waste since 1996. A unique 'feed forward' statistical process control (SPC) has been used rather than statistical quality control (SQC). In SPC, the feed composition to the melter is controlled prior to vitrification. In SQC, the glass product is sampled after it is vitrified. Individual glass property models form the basis for the 'feed forward' SPC. The property models transform constraints on the melt and glass properties into constraints on the feed composition. The property models are mechanistic and depend on glass bonding/structure, thermodynamics, quasicrystalline melt species, and/or electron transfers. The mechanistic models have been validated over composition regions well outside of the regions for which they were developed because they are mechanistic. Mechanistic models allow accurate extension to radioactive and hazardous waste melts well outside the composition boundaries for which they were developed.

This transportation study assumes that defense high-levelwaste is stored in three locations (the Savannah River, Hanford, and Idaho Falls plants) and may be disposed of in (1) a commercial repository or (2) a defense-only repository, either of which could be located at one of the five candidate sites; also documented is a preliminary analysis of the costs and risks of transporting defense high-levelwaste from the three storage sites to the five potential candidate repository sites. 17 references, 4 figures, 27 tables.

Lead-iron phosphate glasses containing a highlevel of Fe.sub.2 O.sub.3 for use as a storage medium for high-level radioactive nuclear waste. By combining lead-iron phosphate glass with various types of simulated high-level nuclear waste, a highly corrosion resistant, homogeneous, easily processed glass can be formed. For corroding solutions at 90.degree. C., with solution pH values in the range between 5 and 9, the corrosion rate of the lead-iron phosphate nuclear waste glass is at least 10.sup.2 to 10.sup.3 times lower than the corrosion rate of a comparable borosilicate nuclear waste glass. The presence of Fe.sub.2 O.sub.3 in forming the lead-iron phosphate glass is critical. Lead-iron phosphate nuclear waste glass can be prepared at temperatures as low as 800.degree. C., since they exhibit very low melt viscosities in the 800.degree. to 1050.degree. C. temperature range. These waste-loaded glasses do not readily devitrify at temperatures as high as 550.degree. C. and are not adversely affected by large doses of gamma radiation in H.sub.2 O at 135.degree. C. The lead-iron phosphate waste glasses can be prepared with minimal modification of the technology developed for processing borosilicate glass nuclear wasteforms.

Hanford high-level liquid waste will be converted into a glass form for long-term storage. The glass must meet certain constraints on its composition and properties in order to have desired properties for processing (e.g., electrical conductivity, viscosity, and liquidus temperature) and acceptable durability for long-term storage. The Optimal Waste Loading (OWL) models, based on rigorous mathematical optimization techniques, have been developed to minimize the number of glass logs required and determine glass-former compositions that will produce a glass meeting all relevant constraints. There is considerable uncertainty in many of the models and data relevant to the formulation of high-level glass. In this paper, we discuss how we handle uncertainty in the glass property models and in the high-levelwaste composition to the vitrification process. Glass property constraints used in optimization are inequalities that relate glass property models obtained by regression analysis of experimental data to numerical limits on property values. Therefore, these constraints are subject to uncertainty. The sampling distributions of the regression models are used to describe the uncertainties associated with the constraints. The optimization then accounts for these uncertainties by requiring the constraints to be satisfied within specified confidence limits. The uncertainty in waste composition is handled using stochastic optimization. Given means and standard deviations of component masses in the high-levelwaste stream, distributions of possible values for each component are generated. A series of optimization runs is performed; the distribution of each waste component is sampled for each run. The resultant distribution of solutions is then statistically summarized. The ability of OWL models to handle these forms of uncertainty make them very useful tools in designing and evaluating high-levelwaste glasses formulations.

The potential health and environmental risk-related impacts of actinide burning for high-levelwaste disposal were evaluated. Actinide burning, also called waste partitioning-transmutation, is an advanced method for radioactive waste management based on the idea of destroying the most toxic components in the waste. It consists of two steps: (1) selective removal of the most toxic radionuclides from high-level/spent fuel waste and (2) conversion of those radionuclides into less toxic radioactive materials and/or stable elements. Risk, as used in this report, is defined as the probability of a failure times its consequence. Actinide burning has two potential health and environmental impacts on waste management. Risks and the magnitude of high-consequence repository failure scenarios are decreased by inventory reduction of the long-term radioactivity in the repository. (What does not exist cannot create risk or uncertainty.) Risk may also be reduced by the changes in the waste characteristics, resulting from selection of waste forms after processing, that are superior to spent fuel and which lower the potential of transport of radionuclides from waste form to accessible environment. There are no negative health or environmental impacts to the repository from actinide burning; however, there may be such impacts elsewhere in the fuel cycle.

This paper describes the chemical performance of the Next-Generation Caustic-Side Solvent Extraction (NG-CSSX) process in its current state of development for removal of cesium from the alkaline high-level tank wastes at the Savannah River Site (SRS) in the US Department of Energy (USDOE) complex. Overall, motivation for seeking a major enhancement in performance for the currently deployed CSSX process stems from needs for accelerating the cleanup schedule and reducing the cost of salt-waste disposition. The primary target of the NG-CSSX development campaign in the past year has been to formulate a solvent system and to design a corresponding flowsheet that boosts the performance of the SRS Modular CSSX Unit (MCU) from a current minimum decontamination factor of 12 to 40,000. The chemical approach entails use of a more soluble calixarene-crown ether, called MaxCalix, allowing the attainment of much higher cesium distribution ratios (DCs) on extraction. Concurrently decreasing the Cs-7SB modifier concentration is anticipated to promote better hydraulics. A new stripping chemistry has been devised using a vitrification-friendly aqueous boric acid strip solution and a guanidine suppressor in the solvent, resulting in sharply decreased DCs on stripping. Results are reported herein on solvent phase behavior and batch Cs distribution for waste simulants and real waste together with a preliminary flowsheet applicable for implementation in the MCU. The new solvent will enable MCU to process a much wider range of salt feeds and thereby extend its service lifetime beyond its design life of three years. Other potential benefits of NG-CSSX include increased throughput of the SRS Salt Waste Processing Facility (SWPF), currently under construction, and an alternative modular near-tank application at Hanford.

The thermal performance of commercial high-levelwaste packages was evaluated on a preliminary basis for the candidate Yucca Mountain repository site. The purpose of this study is to provide an estimate for waste package component temperatures as a function of isolation time in tuff. Several recommendations are made concerning the additional information and modeling needed to evaluate the thermal performance of the Yucca Mountain repository system.

The various national programs for developing and applying technology for the interim storage of spent fuel, high-level radioactive waste, and TRU wastes are summarized. Primary emphasis of the report is on dry storage techniques for uranium dioxide fuels, but data are also provided concerning pool storage.

Sludge-contaminated waste consists of waste contaminated with both insoluble species (the sludge fraction) and entrained supernate. The WCS is based on the assumption that approximately 70% of the weight of what is commonly referred to as sludge is interstitial supernate; the remaining approximately 30% consists of the insoluble species (Reference 1). Development of a method for characterization of sludge-contaminated waste must consider both fractions. Separate waste cuts may contain sludge and supernate fractions in varying proportions due to the nature of the job generating the waste and the variability in waste handling techniques. Development of a distribution representative of all sludge-contaminated waste cuts must allow for varying fractions of sludge and supernate contamination. This document will develop a radionuclide distribution in accordance with the methodology outlined in WSRC 1S SRSWaste Acceptance Criteria Manual, Procedure 2.02, Revision 8 for the sludge fraction of sludge-contaminated waste generated in the F-Area Tank Farm This distribution was based on the assumption that sludge-contaminated waste from F-Area Tank Farm Waste Tanks could be co-mingled, and the actual contamination present on waste in a series of containers from these tanks will be representative of the mean radionuclide distribution. The original characterization was based primarily on process knowledge and fill histories (Reference 6). A single, comprehensive characterization for supernate has been developed previously (Reference 9). This document also describes the methodology for application of radionuclide distributions representative of the sludge and supernate fractions of sludge-contaminated waste to individual waste packages. Most of the waste contaminated with sludge from the F-Area Tank Farm will be categorized as Low Level Waste (LLW) and disposed of in the E-area trenches. The waste does, however, have the potential to be categorized as TRU and/or mixed waste

This report describes recent efforts to develop glass property models that can be used to help estimate the volume of high-levelwaste (HLW) glass that will result from vitrification of Hanford tank waste. The compositions of acceptable and processable HLW glasses need to be optimized to minimize the waste-form volume and, hence, to save cost. A database of properties and associated compositions for simulated waste glasses was collected for developing property-composition models. This database, although not comprehensive, represents a large fraction of data on waste-glass compositions and properties that were available at the time of this report. Glass property-composition models were fit to subsets of the database for several key glass properties. These models apply to a significantly broader composition space than those previously publised. These models should be considered for interim use in calculating properties of Hanford waste glasses.

No public policy issue has been as difficult as high-level nuclear waste. Debates continue regarding Yucca Mountain as a disposal site, and-more generally-the appropriateness of geologic disposal and the need to act quickly. Previous research has focused on possible social, political, and economic consequences of a facility in Nevada. Impacts have been predicted to be potentially large and to emanate mainly from stigmatization of the region due to increased perceptions of risk. Analogous impacts from leaving waste at power plants have been either ignored or assumed to be negligible. This paper presents survey results on attitudes of residents in three counties where nuclear waste is currently stored. Topics include perceived risk, knowledge of nuclear waste and radiation, and impacts on jobs, tourism, and housing values from leaving waste on site. Results are similar to what has been reported for Nevada; the public is concerned about possible adverse effects from on-site storage of waste. PMID:8693158

Disclosed are lead-iron phosphate glasses containing a highlevel of Fe/sub 2/O/sub 3/ for use as a storage medium for high-level radioactive nuclear waste. By combining lead-iron phosphate glass with various types of simulated high-level nuclear waste

There exists a paramount need for improved understanding of the behavior of high-level nuclear waste containers and the impact on structural integrity in terms of leak tightness and mechanical stability. The current program, which at the time of this writing is in its early stage...

The Offices of Energy Research and Environmental Management are immediately concerned with the development of storage/immobilization media for high-level nuclear wastes and excess weapons plutonium. These media must be stable and free of risk to the public or to the environment f...

Structural Integrity of the Hanford High-LevelWaste Tanks were evaluated based on the existing Design and Analysis Documents. All tank structures were found adequate for the normal operating and seismic loads. Potential failure modes of the tanks were assessed by engineering interpretation and extrapolation of the existing engineering documents.

Liquidus temperatures (TL) were measured for high-levelwaste (HLW) borosilicate glasses covering a Savannah River composition region. The primary crystallization phase for most glasses was spinel, a solid solution of trevorite (NiFe2O4) with other oxides (FeO, MnO, and Cr2O3). T...

The research on spinel settling in the high-levelwaste (HLW) glass melter was conducted to assist HLW retrieval, glass formulation, feed preparation, and melter design and operation for minimum-risk and minimum-cost vitrification of HLW at Hanford and other DOE sites. The main r...

The objective is to develop predictive models of organic degradation in high-levelwastes (HLW). We make this information available to facility operators on the Hanford Site to support decision-making processes regarding safety, retrieval, and treatment issues. Emphasis is placed...

The equipment as designed, started, and operated for high-level radioactive waste vitrification at the West Valley Demonstration Project in western New York State is described. Equipment for the processes of melter feed make-up, vitrification, canister handling, and off-gas treatment are included. For each item of equipment the functional requirements, process description, and hardware descriptions are presented.

Nearly half of the highlevel radioactive waste stored at Hanford is composed of highly alkaline concentrates referred to as either salt cakes or Double-Shell Slurry (DSS), depending on their compositions and processing histories. The major components of these concentrates are wa...

The development of an approximate-reasoning (AR)-based model to analyze pretreatment options for high-levelwaste is presented. AR methods are used to emulate the processes used by experts in arriving at a judgment. In this paper, the authors first consider two specific issues in applying AR to the analysis of pretreatment options. They examine how to combine quantitative and qualitative evidence to infer the acceptability of a process result using the example of cesium content in low-level waste. They then demonstrate the use of simple physical models to structure expert elicitation and to produce inferences consistent with a problem involving waste particle size effects.

The technical status of the old U.S. mailine program for highlevel radioactive nuclear waste management, and the newly-developing program for disposal of unreprocessed spent fuel was assessed. The method of long term containment for both of these waste forms is considered to be deep geologic isolation in bedded salt. Each major component of both waste management systems is analyzed in terms of its scientific feasibility, technical achievability and engineering achievability. The resulting matrix leads to a systematic identification of major unresolved technical or scientific questions and/or gaps in these programs.

A workshop on the Role of Natural Analogs in Geologic Disposal of High-Level Nuclear Waste (HLW) was held in San Antonio, Texas, on July 22-25, 1991. It was sponsored by the US Nuclear Regulatory Commission (NRC) and the Center for Nuclear Waste Regulatory Analyses (CNWRA). Invitations to the workshop were extended to a large number of individuals with a variety of technical and professional interests related to geologic disposal of nuclear waste and natural analog studies. The objective of the workshop was to examine the role of natural analog studies in performance assessment, site characterization, and prioritization of research related to geologic disposal of HLW.

Groundwater monitoring at the Savannah River Site (SRS) is required at dozens of waste sites and includes sampling at over 1,000 monitoring wells. The expected longevity of groundwater contamination and associated groundwater monitoring and reporting constitutes a significant long-term cost that represents an increasing proportion of the environmental management budget as surface waste units are closed. Therefore, a comprehensive evaluation of the monitoring program for eighteen regulated waste units was conducted to identify areas where monitoring could be optimized. The units evaluated varied considerably in the scope of monitoring; ranging from two wells to hundreds of wells. In order to systematically evaluate such disparate monitoring networks, SRS developed a decision-logic analysis using flow sheets to address potential areas of optimization. Five areas were identified for evaluation, including: (1) Comparison of current monitoring to regulatory requirements, (2) Spatial distribution, (3) Temporal sampling, (4) Analyte requirements, and (5) Reporting frequency and content. Optimization recommendations were made for fifteen of the eighteen groundwater units. The spatial evaluation resulted in recommendations to suspend sampling in 79 wells and add sampling at 16 wells. The temporal evaluation resulted in recommendations to reduce the number of well visits per year by 504. Analyte reductions were recommended at three groundwater units, with increases at three other units. Reporting frequency reductions were recommended for five units. Approximately $700,000 (direct dollars) of potential annualized cost savings were identified for these groundwater units, provided all recommendations are approved. The largest area of savings was associated with reducing the reporting frequency. The optimization approach has been presented to the EPA and South Carolina Department of Environmental Control (SCHDEC), with unit-specific recommendations approved for all five units

In the last 3 years, significant advances in ceramic technology for high-level nuclear waste solidification have been made. Product quality in terms of leach-resistance, compositional uniformity, structural integrity, and thermal stability promises to be superior to borosilicate glass. This paper addresses the process effectiveness and preliminary designs for glass and ceramic immobilization plants. The reference two-step ceramic process utilizes fluid-bed calcination (FBC) and hot isostatic press (HIP) consolidation. Full-scale demonstration of these well-developed processing steps has been established at DOE and/or commercial facilities for processing radioactive materials. Based on Savannah River-type waste, our model predicts that the capital and operating cost for the solidification of high-level nuclear waste is about the same for the ceramic and glass options. However, when repository costs are included, the ceramic option potentially offers significantly better economics due to its high waste loading and volume reduction. Volume reduction impacts several figures of merit in addition to cost such as system logistics, storage, transportation, and risk. The study concludes that the ceramic product/process has many potential advantages, and rapid deployment of the technology could be realized due to full-scale demonstrations of FBC and HIP technology in radioactive environments. Based on our finding and those of others, the ceramic innovation not only offers a viable backup to the glass reference process but promises to be a viable future option for new high-level nuclear waste management opportunities.

Three copper-based alloys, CDA 102 (oxygen-free copper), CDA 613 (Cu-7Al), and CDA 715 (Cu-30Ni), are being considered along with three austenitic candidates as possible materials for fabrication of containers for disposal of high-level radioactive waste. The waste will include spent fuel assemblies from reactors as well as high-level reprocessing wastes in borosilicate glass and will be sent to the prospective repository at Yucca Mountain, Nevada, for disposal. The containers must maintain mechanical integrity for 50 yr after emplacement to allow for retrieval of waste during the preclosure phase of repository operation. Containment is required to be substantially complete for up to 300 to 1000 yr. During the early period, the containers will be exposed to high temperatures and high gamma radiation fields from the decay of high-levelwaste. The final closure joint will be critical to the integrity of the containers. This volume surveys the available data on the metallurgy of the copper-based candidate alloys and the welding techniques employed to join these materials. The focus of this volume is on the methods applicable to remote-handling procedures in a hot-cell environment with limited possibility of postweld heat treatment. The three copper-based candidates are ranked on the basis of the various closure techniques. On the basis of considerations regarding welding, the following ranking is proposed for the copper-based alloys: CDA 715 (best) > CDA 102 > CDA 613 (worst). 49 refs., 15 figs., 1 tab.

Current guidelines for disposing of high-level radioactive wastes stored in underground tanks at the US Department of Energy`s Hanford Site call for vitrifying high-levelwaste (HLW) in borosilicate glass and disposing the glass canisters in a deep geologic repository. Disposition of the low-level waste (LLW) is yet to be determined, but it will likely be immobilized in a glass matrix and disposed of on site. To lower the radiological risk associated with the LLW form, methods are being developed to separate {sup 90}Sr from the bulk waste material so this isotope can be routed to the HLW stream. A solvent extraction method is being investigated to separate {sup 90}Sr from acid-dissolved Hanford tank wastes. Results of experiments with actual tank waste indicate that this method can be used to achieve separation of {sup 90}Sr from the bulk waste components. Greater than 99% of the {sup 90}Sr was removed from an acidic dissolved sludge solution by extraction with di-tbutylcyclohexano-18-crown-6 in 1-octanol (the SREX process). The major sludge components were not extracted.

There is a fundamental need to continue research aimed at understanding nepheline and spinel crystal formation in high-levelwaste (HLW) glass. Specifically, the formation of nepheline solids (K/NaAlSiO4) during slow cooling of HLW glass can reduce the chemical durability of the glass, which can cause a decrease in the overall durability of the glass waste form. The accumulation of spinel solids ((Fe, Ni, Mn, Zn)(Fe, Cr)2O4), while not detrimental to glass durability, can cause an array of processing problems inside HLW glass melters. In this review, the fundamental differences between glass and solid-crystals are explained using kinetic, thermodynamic, and viscosity arguments, and several highlights of glass-crystallization research, as it pertains to high-levelwaste vitrification, are described. In terms of mitigating spinel in the melter and both spinel and nepheline formation in the canister, the complexity of HLW glass and the intricate interplay between thermal, chemical, and kinetic factors further complicates this understanding. However, new experiments seeking to elucidate the contributing factors of crystal nucleation and growth in waste glass, and the compilation of data from older experiments, may go a long way towards helping to achieve higher waste loadings while developing more efficient processing strategies. Higher waste loadings and more efficient processing strategies will reduce the overall HLW Hanford Tank Waste Treatment and Immobilization Plant (WTP) vitrification facilities mission life.

Hanford Site processes associated with the chemical separation of plutonium from uranium and other fission products produced a variety of volatile, semivolatile, and nonvolatile organic and inorganic waste chemicals that were sent to high-levelwaste tanks. These chemicals have undergone and continue to undergo radiolytic and thermal reactions in the tanks to produce a wide variety of degradation reaction products. The origins of the organic wastes, the chemical reactions they undergo, and their reaction products have recently been examined by Stock (2004). Stock gives particular attention to explaining the presence of various types of volatile and semivolatile organic species identified in headspace air samples. This report complements the Stock report by examining the storage of volatile and semivolatile species in the waste, their transport through any overburden of waste to the tank headspaces, the physical phenomena affecting their concentrations in the headspaces, and their eventual release into the atmosphere above the tanks.

This report presents the results of an experimental study investigating the potential uses for silicon-polymer encapsulation of HighLevel Calcine Waste currently stored within the Idaho Nuclear Technology and Engineering Center (INTEC) at the Idaho National Engineering and Environmental Laboratory (INEEL). The study investigated two different applications of silicon polymer encapsulation. One application uses silicon polymer to produce a waste form suitable for disposal at a HighLevel Radioactive Waste Disposal Facility directly, and the other application encapsulates the calcine material for transportation to an offsite melter for further processing. A simulated waste material from INTEC, called pilot scale calcine, which contained hazardous materials but no radioactive isotopes was used for the study, which was performed at the University of Akron under special arrangement with Orbit Technologies, the originators of the silicon polymer process called Polymer Encapsulation Technology (PET). This document first discusses the PET process, followed by a presentation of past studies involving PET applications to waste problems. Next, the results of an experimental study are presented on encapsulation of the INTEC calcine waste as it applies to transportation or disposal of calcine waste. Results relating to long-term disposal include: 1) a characterization of the pilot calcine waste; 2) Toxicity Characteristic Leaching Procedure (TCLP) testing of an optimum mixture of pilot calcine, polysiloxane and special additives; and, 3) Material Characterization Center testing MCC-1P evaluation of the optimum waste form. Results relating to transportation of the calcine material for a mixture of maximum waste loading include: compressive strength testing, 10-m drop test, melt testing, and a Department of Transportation (DOT) oxidizer test.

This report presents the results of an experimental study investigating the potential uses for silicon-polymer encapsulation of HighLevel Calcine Waste currently stored within the Idaho Nuclear Technology and Engineering Center (INTEC) at the Idaho National Engineering and Environmental Laboratory (INEEL). The study investigated two different applications of silicon polymer encapsulation. One application uses silicon polymer to produce a waste form suitable for disposal at a HighLevel Radioactive Waste Disposal Facility directly, and the other application encapsulates the calcine material for transportation to an offsite melter for further processing. A simulated waste material from INTEC, called pilot scale calcine, which contained hazardous materials but no radioactive isotopes was used for the study, which was performed at the University of Akron under special arrangement with Orbit Technologies, the originators of the silicon polymer process called Polymer Encapsulation Technology (PET). This document first discusses the PET process, followed by a presentation of past studies involving PET applications to waste problems. Next, the results of an experimental study are presented on encapsulation of the INTEC calcine waste as it applies to transportation or disposal of calcine waste. Results relating to long-term disposal include: (1) a characterization of the pilot calcine waste; (2) Toxicity Characteristic Leaching Procedure (TCLP) testing of an optimum mixture of pilot calcine, polysiloxane and special additives; and, (3) Material Characterization Center testing MCC-1P evaluation of the optimum waste form. Results relating to transportation of the calcine material for a mixture of maximum waste loading include: compressive strength testing, 10-m drop test, melt testing, and a Department of Transportation (DOT) oxidizer test.

The Department of Energy plans to vitrify approximately 60,000 metric tons of highlevelwaste sludge from underground storage tanks at the Hanford Nuclear Reservation. To reduce the volume of highlevelwaste requiring treatment, a goal has been set to remove about 90 percent of the aluminum, which comprises nearly 70 percent of the sludge. Aluminum in the form of gibbsite and sodium aluminate can be easily dissolved by washing the waste stream with caustic, but boehmite, which comprises nearly half of the total aluminum, is more resistant to caustic dissolution and requires higher treatment temperatures and hydroxide concentrations. In this work, the dissolution kinetics of aluminum species during caustic leaching of actual Hanford highlevelwaste samples is examined. The experimental results are used to develop a shrinking core model that provides a basis for prediction of dissolution dynamics from known process temperature and hydroxide concentration. This model is further developed to include the effects of particle size polydispersity, which is found to strongly influence the rate of dissolution.

The Nevada Nuclear Waste Storage Investigations (NNWSI) Waste Package project is part of the US Department of Energy`s Civilian Radioactive Waste Management (CRWM) Program. The NNWSI project is working towards the development of multibarriered packages for the disposal of spent fuel and high-levelwaste in tuff in the unsaturated zone at Yucca Mountain at the Nevada Test Site (NTS). The final engineered barrier system design may be composed of a waste form, canister, overpack, borehole liner, packing, and the near field host rock, or some combination thereof. Lawrence Livermore National Laboratory`s (LLNL) role is to design, model, and test the waste package subsystem for the tuff repository. At the present stage of development of the nuclear waste management program at LLNL, the detailed requirements for the waste package design are not yet firmly established. In spite of these uncertainties as to the detailed package requirements, we have begun the conceptual design stage. By conceptual design, we mean design based on our best assessment of present and future regulatory requirements. We anticipate that changes will occur as the detailed requirements for waste package design are finalized. 17 references, 4 figures, 10 tables.

In support of Savannah River Site (SRS) tank closure efforts, the Savannah River National Laboratory (SRNL) conducted Real Waste Testing (RWT) to evaluate Enhanced Chemical Cleaning (ECC), an alternative to the baseline 8 wt% oxalic acid (OA) chemical cleaning technology for tank sludge heel removal. ECC utilizes a more dilute OA solution (2 wt%) and an oxalate destruction technology using ozonolysis with or without the application of ultraviolet (UV) light. SRNL conducted tests of the ECC process using actual SRSwaste material from Tanks 5F and 12H. The previous phase of testing involved testing of all phases of the ECC process (sludge dissolution, OA decomposition, product evaporation, and deposition tank storage) but did not involve the use of UV light in OA decomposition. The new phase of testing documented in this report focused on the use of UV light to assist OA decomposition, but involved only the OA decomposition and deposition tank portions of the process. Compared with the previous testing at analogous conditions without UV light, OA decomposition with the use of UV light generally reduced time required to reach the target of <100 mg/L oxalate. This effect was the most pronounced during the initial part of the decomposition batches, when pH was <4. For the later stages of each OA decomposition batch, the increase in OA decomposition rate with use of the UV light appeared to be minimal. Testing of the deposition tank storage of the ECC product resulted in analogous soluble concentrations regardless of the use or non-use of UV light in the ECC reactor.

This report provides an overview of the international program to study seabed disposal of nuclear wastes. Its purpose is to inform legislators, other policy makers, and the general public as to the history of the program, technological requirements necessary for feasibility assessment, legal questions involved, international coordination of research, national policies, and research and development activities. Each of these major aspects of the program is presented in a separate section. The objective of seabed burial, similar to its continental counterparts, is to contain and to isolate the wastes. The subseabed option should not be confuesed with past practices of ocean dumping which have introduced wastes into ocean waters. Seabed disposal refers to the emplacement of solidified high-level radioactive waste (with or without reprocessing) in certain geologically stable sediments of the deep ocean floor. Specially designed surface ships would transport waste canisters from a port facility to the disposal site. Canisters would be buried from a few tens to a few hundreds of meters below the surface of ocean bottom sediments, and hence would not be in contact with the overlying ocean water. The concept is a multi-barrier approach for disposal. Barriers, including waste form, canister, ad deep ocean sediments, will separate wastes from the ocean environment. High-levelwastes (HLW) would be stabilized by conversion into a leach-resistant solid form such as glass. This solid would be placed inside a metallic canister or other type of package which represents a second barrier. The deep ocean sediments, a third barrier, are discussed in the Feasibility Assessment section. The waste form and canister would provide a barrier for several hundred years, and the sediments would be relied upon as a barrier for thousands of years. 62 references, 3 figures, 2 tables.

A reference design and operational procedures for the disposal of high-level radioactive waste in deep boreholes have been developed and documented. The design and operations are feasible with currently available technology and meet existing safety and anticipated regulatory requirements. Objectives of the reference design include providing a baseline for more detailed technical analyses of system performance and serving as a basis for comparing design alternatives. Numerous factors suggest that deep borehole disposal of high-level radioactive waste is inherently safe. Several lines of evidence indicate that groundwater at depths of several kilometers in continental crystalline basement rocks has long residence times and low velocity. High salinity fluids have limited potential for vertical flow because of density stratification and prevent colloidal transport of radionuclides. Geochemically reducing conditions in the deep subsurface limit the solubility and enhance the retardation of key radionuclides. A non-technical advantage that the deep borehole concept may offer over a repository concept is that of facilitating incremental construction and loading at multiple perhaps regional locations. The disposal borehole would be drilled to a depth of 5,000 m using a telescoping design and would be logged and tested prior to waste emplacement. Waste canisters would be constructed of carbon steel, sealed by welds, and connected into canister strings with high-strength connections. Waste canister strings of about 200 m length would be emplaced in the lower 2,000 m of the fully cased borehole and be separated by bridge and cement plugs. Sealing of the upper part of the borehole would be done with a series of compacted bentonite seals, cement plugs, cement seals, cement plus crushed rock backfill, and bridge plugs. Elements of the reference design meet technical requirements defined in the study. Testing and operational safety assurance requirements are also defined. Overall

United States practice has been to design high-level nuclear waste (HLW) geological repositories with waste densities sufficiently high that repository temperatures surrounding the waste will exceed 100{degrees}C and could reach 250{degrees}C. Basalt and devitrified vitroclastic tuff are among the host rocks considered for waste emplacement. Near-field repository thermal behavior and chemical alteration in such rocks is expected to be similar to that observed in many geothermal systems. Therefore, the predictive modeling required for performance assessment studies of the near field could be validated and calibrated using geothermal systems as natural analogues. Examples are given which demonstrate the need for refinement of the thermodynamic databases used in geochemical modeling of near-field natural analogues and the extent to which present models can predict conditions in geothermal fields.

The Hanford site is located in southeastern Washington State and stores more than 200,000 m3 (55 million gallons) of high-level radioactive waste resulting from the production and processing of plutonium. The waste is stored in large carbon steel tanks that were constructed between 1943 and 1986. The leak and structurally integrity of the more recently constructed double-shell tanks must be maintained until the waste can be removed from the tanks and encapsulated in glass logs for final disposal in a repository. There are a number of corrosion-related threats to the waste tanks, including stress-corrosion cracking, pitting corrosion, and corrosion at the liquid-air interface and in the vapor space. This article summarizes the corrosion management program at Hanford to mitigate these threats.

This report provides a summary of operational experiences, component and system performance, and lessons learned associated with the operation of the Vitrification Facility (VF) at the West Valley Demonstration Project (WVDP). The VF was designed to convert stored high-level radioactive waste (HLW) into a stable waste form (borosilicate glass) suitable for disposal in a federal repository. Following successful completion on nonradioactive test, HLW processing began in July 1995. Completion of Phase 1 of HLW processing was reached on 10 June 1998 and represented the processing of 9.32 million curies of cesium-137 (Cs-137) and strontium-90 (Sr-90) to fill 211 canisters with over 436,000 kilograms of glass. With approximately 85% of the total estimated curie content removed from underground waste storage tanks during Phase 1, subsequent operations will focus on removal of tank heel wastes.

SIA Radon is responsible for management of low- and intermediate-level radioactive waste (LILW) produced in Central Russia. In cooperation with Minatom organizations Radon carries out R and D programs on treatment of simulated highlevelwaste (HLW) as well. Radon scientists deal with a study of materials for LILW, HLW, and Nuclear Power Plants (NPP) wastes immobilization, and development and testing of processes and technologies for waste treatment and disposal. Radon is mostly experienced in LILW vitrification. This experience can be carried over to HLW vitrification especially in field of melting systems. The melter chosen as a basic unit for the vitrification plant is a cold crucible. Later on Radon experience in LILW vitrification as well as our results on simulated HLW vitrification are briefly described.

Modeling free convection heat transfer in an cylindrical annular enclosure is still an active area of research and an important problem to be addressed in the highlevel nuclear waste repository. For the vertically emplaced waste container, the air gap which is between the container shell and the rock borehole, have an important role of dissipating heat to surrounding rack. These waste containers are vertically emplaced in the borehole 300 meters below ground, and in a horizontal grid of 30 {times} 8 meters apart. The borehole will be capped after the container emplacement. The expected initial heat generated is between 3--4.74 kW per container depending on the type of waste. The goal of this study is to use a computer simulation model to find the borehole wall, air-gap and the container outer wall temperature distributions.

This study evaluates alternatives for high-levelwaste forms, containers, container processing systems, and onsite interim storage. Glass waste forms considered are cullet, marbles, gems, and monolithic glass. Small and large containers configured with several combinations of overpack confinement and shield casks are evaluated for these waste forms. Onsite interim storage concepts including canister storage building, bore holes, and storage pad were configured with various glass forms and canister alternatives. All favorable options include the monolithic glass production process as the waste form. Of the favorable options the unshielded 4- and 7-canister overpack options have the greatest technical assurance associated with their design concepts due to their process packaging and storage methods. These canisters are 0.68 m and 0.54 m in diameter respectively and 4.57 m tall. Life-cycle costs are not a discriminating factor in most cases, varying typically less than 15 percent.

Engineers at the Idaho National Engineering and Environmental Laboratory (INEEL) are investigating the use of a proprietary silicon-polymer to encapsulate high-level calcine waste stored at the INEEL's Idaho Nuclear Technology and Engineering Center (INTEC). The silicon-polymer-encapsulated waste may be suitable for direct disposal at a radioactive waste disposal facility or for transport to an offsite melter for further processing. In connection with silicon-polymer encapsulation, the University of Akron, under special arrangement with Orbit Technologies, the originator of the Polymer Encapsulation Technology (PET), has studied a simulated waste material from INTEC called pilot-scale calcine that contains hazardous materials but no radioactive isotopes. In this study, Toxicity Characteristic Leaching Procedure (TCLP) and Materials Characterization Center Test 1P were performed to test the waste form for disposal. In addition, a maximum waste loading was established for transporting the calcine waste at INTEC to an offsite melter. For this maximum waste loading, compressive strength testing, 10-m drop testing, melt testing, and a Department of Transportation (DOT) oxidizer test were performed.

Engineers at the Idaho National Engineering and Environmental Laboratory (INEEL) are investigating the use of a proprietary silicon-polymer to encapsulate high-level calcine waste stored at the INEEL's Idaho Nuclear Technology and Engineering Center (INTEC). The silicon-polymer-encapsulated waste may be suitable for direct disposal at a radioactive waste disposal facility or for transport to an offsite melter for further processing. In connection with silicon-polymer encapsulation, the University of Akron, under special arrangement with Orbit Technologies, the originator of the Polymer Encapsulation Technology (PET), has studied a simulated waste material from INTEC called pilot-scale calcine that contains hazardous materials but no radioactive isotopes. In this study, Toxicity Characteristic Leaching Procedure (TCLP) and Materials Characterization Center Test 1P were performed to test the waste form for disposal. In addition, a maximum waste loading was established for transporting the calcine waste at INTEC to an offsite melter. For this maximum waste loading, compressive strength testing, 10-m drop testing, melt testing, and a Department of Transportation (DOT) oxidizer test were performed.

The Idaho Chemical Processing Plant (ICPP), located at the Idaho National Engineering Laboratory (INEL), has reprocessed irradiated nuclear fuels for the US Department of Energy (DOE) since 1951 to recover uranium, krypton-85, and isolated fission products for interim treatment and immobilization. The acidic radioactive high-level liquid waste (HLLW) is routinely stored in stainless steel tanks and then, since 1963, calcined to form a dry granular solid. The resulting high-levelwaste (HLW) calcine is stored in seismically hardened stainless steel bins that are housed in underground concrete vaults. A research and development program has been established to determine the feasibility of treating ICPP HLW calcine using pyrochemical technology.This technology is described.

Human Factors is the discipline concerned with the acquisition of knowledge about human capabilities and limitations, and the application of such knowledge to the design of systems. This paper discusses the range of human factors issues relevant to high-level radioactive waste (HLRW) management systems and, based on examples from other organizations, presents mechanisms through which to assure application of such expertise in the safe, efficient, and effective management and disposal of high-levelwaste. Additionally, specific attention is directed toward consideration of who might be classified as a human factors specialist, why human factors expertise is critical to the success of the HLRW management system, and determining when human factors specialists should become involved in the design and development process.

The process used to find sites suitable for disposal of highlevel radioactive waste is described. Existing and planned activities for screening successively smaller portions of land to identify suitable candidate sites, and for subsequently selecting one or more of these sites for permanent disposal of radioactive wastes are discussed. Environmental effects of the proposed action, including the anticipate range of field studies to characterize various land areas and reasonable alternative siting strategies, are assessed. The environmental assessment provides the basis for a finding of whether or not implementation of this plan will result in significant environmental impacts.

To prove the existence value of nuclear technology for the world of post cold war, demonstration of safe rad-waste disposal is essential. High-level-waste (HLW) certainly is the key issue to be resolved. To assist a rational and persuasive process on various disposal options, an Analytical Hierarchy Process (AHP) for the decision making of HLW management is presented. The basic theory and rationale are discussed, and applications are shown to illustrate the usefulness of the AHP. The authors wish that the AHP can provide a better direction for the current doomed situations of Taiwan nuclear industry, and to exchange with other countries for sharing experiences on the HLW management.

The objective of this document is to summarize scientific information pertinent to evaluating the extent to which high-levelwaste borosilicate glass corrosion and the associated radionuclide release processes are understood for the range of environmental conditions to which waste glass may be exposed in service. Alteration processes occurring within the bulk of the glass (e.g., devitrification and radiation-induced changes) are discussed insofar as they affect glass corrosion. Volume III contains a bibliography of glass corrosion studies, including studies that are not cited in Volumes I and II.

The objective of this document is to summarize scientific information pertinent to evaluating the extent to which high-levelwaste borosilicate glass corrosion and the associated radionuclide release processes are understood for the range of environmental conditions to which waste glass may be exposed in service. Alteration processes occurring within the bulk of the glass (e.g., devitrification and radiation-induced changes) are discussed insofar as they affect glass corrosion.This document is organized into three volumes. Volumes I and II represent a tiered set of information intended for somewhat different audiences. Volume I is intended to provide an overview of waste glass corrosion, and Volume 11 is intended to provide additional experimental details on experimental factors that influence waste glass corrosion. Volume III contains a bibliography of glass corrosion studies, including studies that are not cited in Volumes I and II. Volume I is intended for managers, decision makers, and modelers, the combined set of Volumes I, II, and III is intended for scientists and engineers working in the field of high-levelwaste.

This technical manual summarizes process and equipment technology developed at Pacific Northwest Laboratory over the last 20 years for vitrification of high-level liquid waste by the Spray Calciner/In-Can Melter process. Pacific Northwest Laboratory experience includes process development and demonstration in laboratory-, pilot-, and full-scale equipment using nonradioactive synthetic wastes. Also, laboratory- and pilot-scale process demonstrations have been conducted using actual high-level radioactive wastes. In the course of process development, more than 26 tonnes of borosilicate glass have been produced in 75 canisters. Four of these canisters contained radioactive waste glass. The associated process and glass chemistry is discussed. Technology areas described include calciner feed treatment and techniques, calcination, vitrification, off-gas treatment, glass containment (the canister), and waste glass chemistry. Areas of optimization and site-specific development that would be needed to adapt this base technology for specific plant application are indicated. A conceptual Spray Calciner/In-Can Melter system design and analyses are provided in the manual to assist prospective users in evaluating the process for plant application, to provide equipment design information, and to supply information for safety analyses and environmental reports. The base (generic) technology for the Spray Calciner/In-Can Melter process has been developed to a point at which it is ready for plant application.

There is a fundamental need to continue research aimed at understanding nepheline and spinel crystal formation in high-levelwaste (HLW) glass. Specifically, the formation of nepheline solids (K/NaAlSiO₄) during slow cooling of HLW glass can reduce the chemical durability of the glass, which can cause a decrease in the overall durability of the glass waste form. The accumulation of spinel solids ((Fe, Ni, Mn, Zn)(Fe,Cr)₂O₄), while not detrimental to glass durability, can cause an array of processing problems inside of HLW glass melters. In this review, the fundamental differences between glass and solid-crystals are explained using kinetic, thermodynamic, and viscosity arguments, and several highlights of glass-crystallization research, as it pertains to high-levelwaste vitrification, are described. In terms of mitigating spinel in the melter and both spinel and nepheline formation in the canister, the complexity of HLW glass and the intricate interplay between thermal, chemical, and kinetic factors further complicates this understanding. However, new experiments seeking to elucidate the contributing factors of crystal nucleation and growth in waste glass, and the compilation of data from older experiments, may go a long way towards helping to achieve higher waste loadings while developing more efficient processing strategies.

Savannah River Technology Center (SRTC) has developed a simpler, faster and more accurate method of analysis for the measurement of silicon in HighLevelWaste (HLW) supernate samples prior to evaporation. The warm acid strike method compared favorably to the current filtration method in the measurement of five samples from three different waste tanks. The technical enhancement of the warm acid strike method is that it will help dissolve small solid particles of silicon-bearing materials that contribute to the scatter in measurements. The method is not intended to measure silicon in samples that have significant quantities of entrained waste tank sludge or DWPF frit particles. Since the method is simple, multiple waste tank samples can be prepared and analyzed at the same time. This improvement should significantly reduce turnaround time for Evaporator Qualification samples at SRTC. Enhanced QC and blind standards have been incorporated into the method for better traceability and overall accuracy. ADS recommends that HighLevelWaste switch to the Acid Strike Methodology using 3 M acid and 4 hours of heating for Evaporator Qualification samples as soon as possible. Extra work is necessary before transfer of the method to C-Lab for routine measurements.

The Failure Modes and Release Quantities of the Hanford HighLevelWaste Tanks due to postulated accident loads were established by a DELPHI Expert Panel consisting of both on-site and off-site experts in the field of Structure and Release. The Report presents the evaluation process, accident loads, tank structural failure conclusion reached by the panel during the two-day meeting.

Igneous rocks of basic composition (basalts, diabases, gabbro-dolerites, dunites, etc.) are an appropriate geological environment for high-levelwaste disposal. During interaction with hot ground waters their isolation ability will increase due to the decrease of hydraulic permeability and increase of their sorption ability. According to petrophysical characteristics, such rocks are viscous-rigid media with the highest mechanical stability and do not undergo any changes in properties over the whole temperature range, which is possible in a HLW repository.

Westinghouse Idaho Nuclear Company (WINCO) is specifying and designing a new highlevelwaste tank farm at the Idaho National Engineering Laboratory (INEL). The farm consists of four underground storage tanks, which replace the existing tanks. The new facility includes provisions for remote operations. One of the planned remote operations is robotic inspection of the tank from the interior and exterior. This paper describes the process used to design the robotic system for the inspection tasks.

A suite of multi-dimensional computer models was developed in 1999 (Collard and Flach) to analyze the transport of residual contamination from high-levelwaste tanks through the subsurface to seeplines. Enhancements in 2000 to those models include investigate the effect of numerical dispersion, develop a solubility-limited case for U and Pu, and develop a plan for a database as part of the Rapid Screening Tool and start to implement that plan.

This Safety Analysis Report describes the design, analyses, and principle features of the Vitrified HighLevelWaste (VHLW) Cask. In preparing this report a detailed evaluation of the design has been performed to ensure that all safety, licensing, and operational goals for the cask and its associated Department of Energy program can be met. The functions of this report are: (1) to fully document that all functional and regulatory requirements of 10CFR71 can be met by the package; and (2) to document the design and analyses of the cask for review by the Nuclear Regulatory Commission. The VHLW Cask is the reusable shipping package designed by GNSI under Department of Energy contract DE-AC04-89AL53-689 for transportation of Vitrified HighLevelWaste, and to meet the requirements for certification under 10CFR71 for a Type B(U) package. The VHLW cask has been designed as packaging for transport of canisters of Vitrified HighLevelWaste solidified at Department of Energy facilities.

This report describes the major parameters for management of spent nuclear fuel and high-level radioactive wastes in selected foreign countries as of December 1989 and compares them with those in the United States. The foreign countries included in this study are Belgium, Canada, France, the Federal Republic of Germany, Japan, Sweden, Switzerland, and the United Kingdom. All the countries are planning for disposal of spent fuel and/or high-levelwastes in deep geologic repositories. Most countries (except Canada and Sweden) plan to reprocess their spent fuel and vitrify the resultant high-level liquid wastes; in comparison, the US plans direct disposal of spent fuel. The US is planning to use a container for spent fuel as the primary engineered barrier. The US has the most developed repository concept and has one of the earliest scheduled repository startup dates. The repository environment presently being considered in the US is unique, being located in tuff above the water table. The US also has the most prescriptive regulations and performance requirements for the repository system and its components. 135 refs., 8 tabs.

Three copper-based alloys and three iron- to nickel-based austenitic alloys are being considered as possible materials for fabrication of high-level radioactive-waste disposal containers. The waste will include spent fuel assemblies from reactors as well as high-levelwaste in borosilicate glass and will be sent to the prospective site at Yucca Mountain, Nevada, for disposal. The copper-based alloy materials are CDA 102 (oxygen-free copper), CDA 613 (Cu-7Al), and CDA 715 (Cu-30Ni). The austenitic materials are Types 304L and 316L stainless steels and Alloy 825. The waste-package containers must maintain substantially complete containment for at least 300 yr and perhaps as long as 1000 yr, and they must be retrievable from the disposal site during the first 50 yr after emplacement. The containers will be exposed to high temperatures and high gamma radiation fields from the decay of high-levelwaste. This volume surveys the available data on the phase stability of both groups of candidate alloys. The austenitic alloys are reviewed in terms of the physical metallurgy of the iron-chromium-nickel system, martensite transformations, carbide formation, and intermetallic-phase precipitation. The copper-based alloys are reviewed in terms of their phase equilibria and the possibility of precipitation of the minor alloying constituents. For the austenitic materials, the ranking based on phase stability is: Alloy 825 (best), Type 316L stainless steel, and then Type 304L stainless steel (worst). For the copper-based materials, the ranking is: CDA 102 (oxygen-free copper) (best), and then both CDA 715 and CDA 613. 75 refs., 24 figs., 6 tabs.

The cold crucible inductive melting (CCIM) technology successfully applied for vitrification of low- and intermediate-level waste (LILW) at SIA Radon, Russia, was tested to be implemented for vitrification of high-levelwaste (HLW) stored at Savannah River Site, USA. Mixtures of Sludge Batch 2 (SB2) and 4 (SB4) waste surrogates and borosilicate frits as slurries were vitrified in bench- (236 mm inner diameter) and full-scale (418 mm inner diameter) cold crucibles. Various process conditions were tested and major process variables were determined. Melts were poured into 10L canisters and cooled to room temperature in air or in heat-insulated boxes by a regime similar to Canister Centerline Cooling (CCC) used at DWPF. The products with waste loading from ~40 to ~65 wt.% were investigated in details. The products contained 40 to 55 wt.% waste oxides were predominantly amorphous; at higher waste loadings (WL) spinel structure phases and nepheline were present. Normalized release values for Li, B, Na, and Si determined by PCT procedure remain lower than those from EA glass at waste loadings of up to 60 wt.%.

In 1982, after years of deliberation and in response to mounting pressures from environmental, industrial, and other groups, the US Congress enacted the Nuclear Waste Policy Act (NWPA) of 1982, which was signed into law by the President in January 1983. That legislation signified a major milestone in the nation's management of high-level nuclear waste, since it represented a consensus among the nation's lawmakers to tackle a problem that had evaded solution for decades. Implementation of the NWPA has proven to be exceedingly difficult, as attested by the discord generated by the US Department of Energy's (DOE's) geologic repository and monitored retrievable storage (MRS) facility siting activities. The vision that motivated the crafters of the 1982 act became blurred as opposition to the law increased. After many hearings that underscored the public's concern with the waste management program, the Congress enacted the Nuclear Waste Policy Amendments Act of 1987 (Amendments Act), which steamlined and focused the program, while establishing three independent bodies: the MRS Review Commission, the Nuclear Waste Technical Review Board, and the Office of the Nuclear Waste Negotiator. Yet, even as the program evolves, several themes characterizing the nation's effort to solve the waste management problem continue to prevail. The first of these themes has to do with social consciousness, and the others that follow deal with technical leadership, public involvement and risk perceptions, and program conservatism.

Hanford Site single-shell tank (SSI) and double-shell tank (DSI) wastes are planned to be separated into low activity (or low-level waste, LLW) and high activity (or high-levelwaste, HLW) fractions, and to be vitrified for disposal. Formulation of HLW glass must comply with glass processibility and durability requirements, including constraints on melt viscosity, electrical conductivity, liquidus temperature, tendency for phase segregation on the molten glass surface, and chemical durability of the final waste form. A wide variety of HLW compositions are expected to be vitrified. In addition these wastes will likely vary in composition from current estimates. High concentrations of certain troublesome components, such as sulfate, phosphate, and chrome, raise concerns about their potential hinderance to the waste vitrification process. For example, phosphate segregation in the cold cap (the layer of feed on top of the glass melt) in a Joule-heated melter may inhibit the melting process (Bunnell, 1988). This has been reported during a pilot-scale ceramic melter run, PSCM-19, (Perez, 1985). Molten salt segregation of either sulfate or chromate is also hazardous to the waste vitrification process. Excessive (Cr, Fe, Mn, Ni) spinel crystal formation in molten glass can also be detrimental to melter operation.

The national program for long-term management of highlevelwaste (HLW) from nuclear fuels reprocessing is discussed. This covers only DOE defense wastes. Current emphasis is on solidification of waste into a form that, along with additional barriers, may be permanently stored in a repository. An integrated national plan incorporates all the elements of such an overall HLW disposal system. Interim storage is in near-surface tanks at the Hanford and Savannah River sites. At the Idaho site, waste is stored in bins after being calcined. Some Idaho waste is liquid and is also stored in tanks before calcination. Retrieval and immobilization of HLW into a solid, low-release form represent the major elements for which our long-term program has responsibility. Once solidified, the waste will temporarily remain onsite until the final disposal site is prepared for receipt of waste. Currently, a geologic repository is favored for ultimate disposal, although other possibilities such as seabed, icecap, space, and near-surface disposal are also being considered.

This document reports the results of a study aimed at determining the best strategy for providing containment during management of commercial high-level radioactive wastes. Containment to assure public and worker safety is needed for all storage, transport, handling, and disposal operations. There are several thousand containment system options; this work determined, in overview rather than detail, which options should be pursued. This work shows that the geologic and engineered barriers in repositories in different geologic media, such as salt and granite, play very different roles in preserving long-term containment. In sum, there is no common engineered waste package that is suitable for disposal in all geologic media, each package must be tailored to the specific repository system. The need to make the waste package specific to the repository system leads to the key elements of waste management containment strategy: perform final packaging at the disposal site, and deliver to the site a waste that is in a form suitable for disposal and in a container that is (a) appropriate for the process that produced the waste form, (b) satisfactory for transport, and (c) suitable as the common basis for custom tailoring the waste package for any repository system. As described in this report, mild carbon steel is a container material that can be expected to meet these requirements.

A stress corrosion cracking (SCC) model has been adapted for performance prediction of highlevel radioactive-waste packages to be emplaced in the proposed Yucca Mountain radioactive-waste repository. SCC is one form of environmentally assisted cracking resulting from the presence of three factors: metallurgical susceptibility, critical environment, and tensile stresses. For waste packages of the proposed Yucca Mountain repository, the outer barrier material is the highly corrosion-resistant Alloy UNS-N06022, the environment is represented by the water film present on the surface of the waste package from dripping or deliquescence of soluble salts present in any surface deposits, and the stress is principally the weld induced residual stress. SCC has historically been separated into 'initiation' and 'propagation' phases. Initiation of SCC will not occur on a smooth surface if the surface stress is below a threshold value defined as the threshold stress. Cracks can also initiate at and propagate from flaws (or defects) resulting from manufacturing processes (such as welding). To account for crack propagation, the slip dissolution/film rupture (SDFR) model is adopted to provide mathematical formulae for prediction of the crack growth rate. Once the crack growth rate at an initiated SCC is determined, it can be used by the performance assessment (not in the scope of this paper) to determine the time to through-wall penetration for the waste package. This paper presents the development and validation of the SDFR crack growth rate model based on technical information in the literature as well as experimentally determined crack growth rates developed specifically for Alloy UNS- N06022 in environments relevant to highlevel radioactive-waste packages of the proposed Yucca Mountain radioactive-waste repository.

Characterization of HLW is a prerequisite for effective planning of HLW disposition and site closure performance assessment activities. Adequate characterization typically requires application of a combination of data sources, including process knowledge, theoretical relationships, and real-waste analytical data. Consistently obtaining high quality real-waste analytical data is a challenge, particularly for HLW sludge insolubles, due to the inherent complexities associated with matrix heterogeneities, sampling access limitations, radiological constraints, analyte loss mechanisms, and analyte measurement interferences. Understanding how each of these complexities affects the analytical results is the first step to developing a sampling and analysis program that provides characterization data that are both meaningful and adequate. A summary of the key elements impacting SRS HLW sludge analytical data uncertainties is presented in this paper, along with guidelines for managing each of the impacts. The particular elements addressed include: (a) sample representativeness; (b) solid/liquid phase quantification effectiveness; (c) solids dissolution effectiveness; (d) analyte cross contamination, loss, and tracking; (e) dilution requirements; (f) interference removal; (g) analyte measurement technique; and (h) analytical detection limit constraints. A primary goal of understanding these elements is to provide a basis for quantifying total propagated data uncertainty.

This paper summarizes preliminary remote handling and robotic concepts being developed as part of the US Department of Energy's (DOE) Yucca Mountain Project. The DOE is currently evaluating the Yucca Mountain Nevada site for suitability as a possible underground geologic repository for the disposal of highlevel nuclear waste. The current advanced conceptual design calls for the disposal of more than 12,000 highlevel nuclear waste packages within a 225 km underground network of tunnels and emplacement drifts. Many of the waste packages may weigh as much as 66 tonnes and measure 1.8 m in diameter and 5.6 m long. The waste packages will emit significant levels of radiation and heat. Therefore, remote handling is a cornerstone of the repository design and operating concepts. This paper discusses potential applications areas for robotics and remote handling technologies within the subsurface repository. It also summarizes the findings of a preliminary technology survey which reviewed available robotic and remote handling technologies developed within the nuclear, mining, rail and industrial robotics and automation industries, and at national laboratories, universities, and related research institutions and government agencies.

With the imminent startup, in the United States, of facilities for vitrification of high-level nuclear waste, a document has been prepared that compiles the scientific basis for understanding the alteration of the waste glass products under the range of service conditions to which they may be exposed during storage, transportation, and eventual geologic disposal. A summary of selected parts of the content of this document is provided. Waste glass alterations in a geologic repository may include corrosion of the glass network due to groundwater and/or water vapor contact. Experimental testing results are described and interpreted in terms of the underlying chemical reactions and physical processes involved. The status of mechanistic modeling, which can be used for long-term predictions, is described and the remaining uncertainties associated with long-term simulations are summarized.

With the imminent startup, in the United States, of facilities for vitrification of high-level nuclear waste, a document has been prepared that compiles the scientific basis for understanding the alteration of the waste glass products under the range of service conditions to which they may be exposed during storage, transportation, and eventual geologic disposal. A summary of selected parts of the content of this document is provided. Waste glass alterations in a geologic repository may include corrosion of the glass network due to groundwater and/or water vapor contact. Experimental testing results are described and interpreted in terms of the underlying chemical reactions and physical processes involved. The status of mechanistic modeling, which can be used for long-term predictions, is described and the remaining uncertainties associated with long-term simulations are summarized.

The US Department of Energy is responsible for the design, construction, operation and closure of a repository in which to permanently dispose of the nation`s highlevel nuclear waste. In addition to the objective of safely isolating the waste inventory, the repository must provide a safe working environment for its workforce, and protect the public. The conceptual design for this facility is currently being developed. Tunnel Boring Machine will be used to excavate 228 kilometers of tunneling to construct the facility over a 30 year period. The excavation operations will be physically separated from the waste emplacement operations, and each operation will have its own dedicated ventilation system. The facility is being designed to remain open for 150 years.

The U.S. government has the obligation of managing the high-level radioactive waste from its defense activities and also, under existing law, from civilian nuclear power generation. This obligation is not being met. The January 2012 Final Report from the Blue Ribbon Commission on America's Nuclear Future provides commendable guidance but little that is new. The author, who served on the federal Nuclear Waste Technical Review Board from 1989 to 1994 and subsequently on the Board on Radioactive Waste Management of the National Research Council from 1994 to 1999, provides a perspective both on the Commission's recommendations and a potential path toward progress in meeting the federal obligation. By analogy to Sisyphus of Greek mythology, our nation needs to find a way to roll the rock to the top of the hill and have it stay there, rather than continuing to roll back down again. PMID:23311528

Savannah River Remediation, LLC (SRR) operates the Defense Waste Processing Facility for the U.S. Department of Energy at the Savannah River Site. This facility immobilizes high-level radioactive waste through vitrification following chemical pretreatment. Catalytic destruction of formate and oxalate ions to carbon dioxide has been observed during qualification testing of non-radioactive analog systems. Carbon dioxide production greatly exceeded hydrogen production, indicating the occurrence of a process other than the catalytic decomposition of formic acid. Statistical modeling was used to relate the new reaction chemistry to partial catalytic wet air oxidation of both formate and oxalate ions driven by the low concentrations of palladium, rhodium, and/or ruthenium in the waste. Variations in process conditions led to increases or decreases in the total oxidative destruction, as well as partially shifting the preferred species undergoing destruction from oxalate ion to formate ion.

A geological repository has been proposed for the permanent disposal of the nation`s highlevel nuclear waste at Yucca Mountain in the Nevada desert. The containers for this waste must remain intact for the unprecedented service lifetime of 1000 years. A combination of engineering, regulatory, and licensing requirements complicate the container material selection. In parallel to gathering information regarding the Yucca Mountain service environment and material performance data, a set of selection criteria have been established which compare candidate materials to the performance requirements, and allow a quantitative comparison of candidates. These criteria assign relative weighting to varied topic areas such as mechanical properties, corrosion resistance, fabricability, and cost. Considering the long service life of the waste containers, it is not surprising that the corrosion behavior of the material is a dominant factor. 7 refs.

... Falls, ID 83415, telephone (208) 526-5793. Correction In the Federal Register of January 4, 2010, in FR... Record of Decision: Idaho High-LevelWaste and Facilities Disposition Final Environmental Impact..., announcing an amended Record of Decision (ROD) for the Idaho High-LevelWaste and Facilities...

A stress corrosion cracking (SCC) model has been adapted for performance prediction of highlevel radioactive-waste packages to be emplaced in the proposed Yucca Mountain repository. For waste packages of the proposed Yucca Mountain repository, the outer barrier material is the highly corrosion-resistant Alloy UNS-N06022 (Alloy 22), the environment is represented by aqueous brine films present on the surface of the waste package from dripping or deliquescence of soluble salts present in any surface deposits, and the tensile stress is principally from weld induced residual stress. SCC has historically been separated into ''initiation'' and ''propagation'' phases. Initiation of SCC will not occur on a smooth surface if the surface stress is below a threshold value defined as the threshold stress. Cracks can also initiate at and propagate from flaws (or defects) resulting from manufacturing processes (such as welding); or that develop from corrosion processes such as pitting or dissolution of inclusions. To account for crack propagation, the slip dissolution/film rupture (SDFR) model is adopted to provide mathematical formulae for prediction of the crack growth rate. Once the crack growth rate at an initiated SCC is determined, it can be used by the performance assessment to determine the time to through-wall penetration for the waste package. This paper presents the development of the SDFR crack growth rate model based on technical information in the literature as well as experimentally determined crack growth rates developed specifically for Alloy UNS-N06022 in environments relevant to highlevel radioactive-waste packages of the proposed Yucca Mountain radioactive-waste repository. In addition, a seismic damage related SCC crack opening area density model is briefly described.

Three copper-based alloys and three iron- to nickel-based austenitic alloys are being considered as possible materials for fabrication of containers for disposal of high-level radioactive waste. This waste will include spent fuel assemblies from reactors as well as high-levelwaste in borosilicate glass and will be sent to the prospective site at Yucca Mountain, Nevada, for disposal. The containers must maintain substantially complete containment for at least 300 yr and perhaps as long as 1000 yr. During the first 50 yr after emplacement, they must be retrievable from the disposal site. Shortly after the containers are emplaced in the repository, they will be exposed to high temperatures and high gamma radiation fields from the decay of the high-levelwaste. This volume surveys the available data on oxidation and corrosion of the iron- to nickel-based austenitic materials (Types 304L and 316L stainless steels and Alloy 825) and the copper-based alloy materials (CDA 102 (oxygen-free copper), CDA 613 (Cu-7Al), and CDA 715 (Cu-30Ni)), which are the present candidates for fabrication of the containers. Studies that provided a large amount of data are highlighted, and those areas in which little data exists are identified. Examples of successful applications of these materials are given. On the basis of resistance to oxidation and general corrosion, the austenitic materials are ranked as follows: Alloy 825 (best), Type 316L stainless steel, and then Type 304L stainless steel (worst). For the copper-based materials, the ranking is as follows: CDA 715 and CDA 613 (both best), and CDA 102 (worst). 110 refs., 30 figs., 13 tabs.

High-level nuclear waste is being vitrified, i.e., converted to a durable glass that can be stored in a safe repository for hundreds of thousands of years. Waste vitrification is accomplished in reactors called melters to which the waste is charged together with glass-forming additives. The mixture is electrically heated to a temperature as high as 1150?C (or even higher in advanced melters) to create a melt that becomes glass on cooling. This process is slow and expensive. Moreover, the melters that are currently in use or are going to be used in the U.S. are sensitive to clogging and thus cannot process melt in which solid particles are suspended. These particles settle and gradually accumulate on the melter bottom. Such particles, most often small crystals of spinel (a mineral containing iron, nickel, chromium, and other minor oxides), inevitably occur in the melt when the content of the waste in the glass (called waste loading) increases above a certain limit. To avoid the presence of solid particles in the melter, the waste loading is kept rather low, in average 15% lower than in glass formulated for more robust melters.

High-level nuclear waste is being vitrified, i.e., converted to a durable glass that can be stored in a safe repository for hundreds of thousands of years. Waste vitrification is accomplished in reactors call melters to which the waste is charged together with glass-forming additives. The mixture is electrically heated to a temperature as high as 1150 degree C (or even higher in advanced melters) to create a melt that becomes glass on cooling. This process is slow and expensive. Moreover, the melters that are currently in use or are going to be used in the U.S. are sensitive to clogging and thus cannot process melt in which solid particles are suspended. These particles settle and gradually accumulate on the melter bottom. Such particles, most often small crystals of spinel ( a mineral containing iron, nickel, chromium, and other minor oxides), inevitably occurred in the melt when the content of the waste in the glass (called waste loading) increases above a certain limit. To avoid the presence of solid particles in the melter, the waste loading is kept rather low, in average 15% lower than in glass formulated for more robust melters.

No public policy issue has been as difficult as high-level nuclear waste. Debates continue regarding Yucca Mountain as a disposal site, and - more generally - the appropriateness of geologic disposal and the need to act quickly. Previous research has focused on possible social, political, and economic consequences of a facility in Nevada. Impacts have been predicted to be potentially large and to emanate mainly from stigmatization of the region due to increased perceptions of risk. Analogous impacts from leaving waste at power plants have been either ignored or assumed to be negligible. This paper presents survey results on attitudes of residents in three countries where nuclear waste is currently stored. Topics include perceived risk, knowledge of nuclear waste and radiation, and impacts on jobs, tourism, and housing values from leaving waste on site. Results are similar to what has been reported for Nevada; the public is concerned about possible adverse effects from on-site storage of waste. 24 refs., 7 figs., 5 tabs.

The seven candidate waste forms, evaluated as potential media for the immobilization and gelogic disposal of high-level nuclear wastes were borosilicate glass, SYNROC, tailored ceramic, high-silica glass, FUETAP concrete, coated sol-gel particles, and glass marbles in a lead matrix. The evaluation, completed on August 1, 1981, combined preliminary waste form evaluations conducted at Department of Energy (DOE) defense waste-sites and at independent laboratories, peer review assessments, a product performance evaluation, and a processability analysis. Based on the combined results of these four inputs, two of the seven forms, borosilicate glass and a titanate-based ceramic, SYNROC, were selected as the reference and alternative forms, respectively, for continued development and evaluation in the National HLW Program. The borosilicate glass and ceramic forms were further compared during FY-1982 on the basis of risk assessments, cost comparisons, properties comparisons, and conformance with proposed regulatory and repository criteria. Both the glass and ceramic forms are viable candidates for use at DOE defense HLW sites; they are also candidates for immobilization of commercial reprocessing wastes. This paper describes the waste form screening process, discusses each of the four major inputs considered in the selection of the two forms in 1981, and presents a brief summary of the comparisons of the two forms during 1982 and the selection process to determine the final form for SRP defense HLW.

This report summarizes the work of the Engineering Studies Task Group (ESTG) of the Seabed Working Group during its study of emplacement systems for the subseabed disposal of highlevel radioactive waste. ESTG has performed design studies of emplacement systems, costed them, and estimated operational reliabilities. Mathematical models for important physical and engineering processes were developed and a large number of laboratory tests, sea trials, and in situ experiments for the purpose of understanding the emplacement environment and developing the specialized equipment necessary for emplacement were performed. Attention was focused on two systems. The first would emplace a 450-m column of waste packages in predrilled holes 750 m deep. The second would use free falling gravity penetrators launched from a disposal ship to embed waste packages about 50 m below the seafloor in an array that separated each by an average of 180 m from its neighbors. Studies of each system covered all aspects, from the configuration and functions of the port facilities through transport to the ocean site, emplacement operations, and post emplacement behavior of the waste packages. Cost and reliability studies were similarly broad. ESTG concludes that viable disposal systems for subseabed emplacement of waste are feasible. If appropriate sites can be found, it appears that straightforward methods are available for producing satisfactory waste packages that can survive a 500-yr emplacement period. 172 refs., 40 figs., 19 tabs.

Current plans call for the United States Department of Energy (DOE) to start up facilities for vitrification of high-level radioactive waste (HLW) stored in tanks at the Savannah River Site, Aiken, South Carolina, in 1995; West Valley Demonstration Project, West Valley, New York, in 1996; and at the Hanford Site, Richland, Washington, after the year 2000. The product from these facilities will be canistered HLW borosilicate glass, which will be stored, transported, and eventually disposed of in a geologic repository. The behavior of this glass waste product, under the range of likely service conditions, is the subject of considerable scientific and public interest. Over the past few decades, a large body of scientific information on borosilicate waste glass has been generated worldwide. The intent of this document is to consolidate information pertaining to our current understanding of waste glass corrosion behavior and radionuclide release. The objective, scope, and organization of the document are discussed in Section 1.1, and an overview of borosilicate glass corrosion is provided in Section 1.2. The history of glass as a waste form and the international experience with waste glass are summarized in Sections 1.3 and 1.4, respectively.

The Savannah River Site (SRS) is a Department of Energy site which has produced nuclear materials for national defense, research, space, and medical programs since the 1950's. As a by-product of this activity, approximately 37 million gallons of high-level liquid waste containing approximately 292 million curies of radioactivity is stored on an interim basis in 45 underground storage tanks. Originally, 51 tanks were constructed and utilized to support the mission. Four tanks have been closed and taken out of service and two are currently undergoing the closure process. The Liquid Waste System is a highly integrated operation involving safely storing liquid waste in underground storage tanks; removing, treating, and dispositioning the low-level waste fraction in grout; vitrifying the higher activity waste at the Defense Waste Processing Facility; and storing the vitrified waste in stainless steel canisters until permanent disposition. After waste removal and processing, the storage and processing facilities are decontaminated and closed. A Liquid Waste System Plan (hereinafter referred to as the Plan) was developed to integrate and document the activities required to disposition legacy and future High-LevelWaste and to remove from service radioactive liquid waste tanks and facilities. It establishes and records a planning basis for waste processing in the liquid waste system through the end of the program mission. The integrated Plan which recognizes the challenges of constrained funding provides a path forward to complete the liquid waste mission within all regulatory and legal requirements. The overarching objective of the Plan is to meet all Federal Facility Agreement and Site Treatment Plan regulatory commitments on or ahead of schedule while preserving as much life cycle acceleration as possible through incorporation of numerous cost savings initiatives, elimination of non-essential scope, and deferral of other scope not on the critical path to compliance

This paper provides a brief overview of current repository and engineered barrier system (EBS) designs in selected high-levelwaste (HLW) and spent fuel (SF) disposal concepts from European countries, with special reference to key metallic waste containers and corrosion processes. The paper discusses assessments of copper, iron and steel container corrosion behaviour under the environmental conditions expected, given likely repository host rocks and groundwaters, and comments on the significance of corrosion processes, the choice of waste container materials, and areas of research. Most of the HLW and/or SF disposal programmes in European countries are pursuing disposal options in which the primary waste container is designed, in conjunction with the surrounding EBS materials, to provide complete containment of the waste for at least the period when temperatures in the disposal system are significantly raised by radioactive decay.

Results of the waste form evaluation are summarized as: risks of human exposure are comparable and extremely small for either borosilicate glass or Synroc ceramic. Waste form properties are more than adequate for either form. The waste form decision can therefore be made on the basis of practicality and cost effectiveness. Synroc offers lower costs for transportation and emplacement. The borosilicate glass form offers the lowest total disposal cost, much simpler and less costly production, an established and proven process, lower future development costs, and an earlier startup of the DWPF.

Processes for the removal of residual sludge from SRSwaste tanks have historically used solutions containing up to 0.9 M oxalic acid to dissolve the remaining material following sludge removal. The selection of this process was based on a comparison of a number of studies performed to evaluate the dissolution of residual sludge. In contrast, the dissolution of the actinide mass, which represents a very small fraction of the waste, has not been extensively studied. The Pu, Np, and Am in the sludge is reported to be present as hydrated and crystalline oxides. To identify aqueous solutions which have the potential to increase the solubility of the actinides, the alkaline and mildly acidic test solutions shown below were selected as candidates for use in a series of solubility experiments. The efficiency of the solutions in solubilizing the actinides was evaluated using a simulated sludge prepared by neutralizing a HNO{sub 3} solution containing Pu, Np, and Am. The hydroxide concentration was adjusted to a 1.2 M excess and the solids were allowed to age for several weeks prior to starting the experiments. The sludge was washed with 0.01 M NaOH to prepare the solids for use. Following the addition of an equal portion of the solids to each test solution, the concentrations of Pu, Np, and Am were measured as a function of time over a 792 h (33 day) period to provide a direct comparison of the efficiency of each solution in solubilizing the actinide elements. Although the composition of the sludge was limited to the hydrated actinide oxides (and did not contain other components of demonstrated importance), the results of the study provides guidance for the selection of solutions which should be evaluated in subsequent tests with a more realistic surrogate sludge and actual tank waste.

The Southern Routes for High-Level Radioactive Waste: Agencies, Contacts and Designations is a compendium of sixteen southern states` routing programs for the transportation of high-level radioactive materials. The report identifies the state-designated routing agencies as defined under 49 Code of Federal Regulations (CFR) Part 171 and provides a reference to the source and scope of the agencies` rulemaking authority. Additionally, the state agency and contact designated by the state`s governor to receive advance notification and shipment routing information under 10 CFR Parts 71 and 73 are also listed. This report also examines alternative route designations made by southern states and the lessons that were learned from the designation process.

The Southern Routes for High-Level Radioactive Waste: Agencies, Contacts and Designations is a compendium of sixteen southern states' routing programs for the transportation of high-level radioactive materials. The report identifies the state-designated routing agencies as defined under 49 Code of Federal Regulations (CFR) Part 171 and provides a reference to the source and scope of the agencies' rulemaking authority. Additionally, the state agency and contact designated by the state's governor to receive advance notification and shipment routing information under 10 CFR Parts 71 and 73 are also listed. This report also examines alternative route designations made by southern states and the lessons that were learned from the designation process.

Spent fuels from once-through fuel cycles placed in underground repositories have the potential to become attractive targets for diversion and/or theft because of their valuable material content and decreasing radioactivity. The first geologic repository in the US, as currently designed, will contain approximately 500 Mt of plutonium, 60,000 Mt of uranium and a host of other fissile and strategically important elements. This paper identifies some of the international safeguards issues relevant to the various proposed scenarios for disposing of the spent fuel. In the context of the US program for geologic disposal of spent fuels, this paper highlights several issues that should be addressed in the near term by US industries, the Department of Energy, and the Nuclear Regulatory Commission before the geologic repositories for spent fuels become a reality. Based on US spent fuel discharges, an example is presented to illustrate the enormity of the problem of verifying spent fuel inventories. The geologic disposal scenario for high-levelwastes originating from defense facilities produced a practicably irrecoverable'' waste form. Therefore, safeguards issues for geologic disposal of high-levelwaste now in the US are less pressing. 56 refs. , 2 figs.

A repository located at Yucca Mountain at the Nevada Test Site is a potential site for permanent geological disposal of high-level nuclear waste. The repository can be located in a horizon in welded tuff, a volcanic rock, which is above the static water level at this site. The environmental conditions in this unsaturated zone are expected to be air and water vapor dominated for much of the containment period. Type 304L stainless steel is the reference material for fabricating canisters to contain the solid high-levelwastes. Alternative stainless alloys are considered because of possible susceptibility of 304L to localized and stress forms of corrosion. For the reprocessed glass wastes, the canisters serve as the recipient for pouring the glass with the result that a sensitized microstructure may develop because of the times at elevated temperatures. Corrosion testing of the reference and alternative materials has begun in tuff-conditioned water and steam environments. 21 references, 8 figures, 8 tables.

The Analytical Chemistry Laboratory at the Pacific Northwest Laboratory in Richland, Washington, is actively involved in performing analytical characterization of high-level mixed waste from Hanford`s single shell and double shell tank characterization programs. A full suite of analyses is typically performed on homogenized tank core samples. These analytical techniques include inductively-coupled plasma-atomic emission spectroscopy, total organic carbon methods and radiochemistry methods, as well as many others, all requiring some type of remote sample-preparation treatment to solubilize the tank sludge material for analysis. Most of these analytical methods typically use a single sample-preparation treatment, inherently providing elemental information only. To better understand and interpret tank chemistry and assist in identifying chemical compounds, selected analytical methods are performed using multiple sample-preparation treatments. The sample preparation treatments used at Pacific Northwest Laboratory for this work with high-level mixed waste include caustic fusion, acid digestion, and water leach. The type of information available by comparing results from different sample-prep treatments includes evidence for the presence of refractory compounds, acid-soluble compounds, or water-soluble compounds. Problems unique to the analysis of Hanford tank wastes are discussed. Selected results from the Hanford single shell ferrocyanide tank, 241-C-109, are presented, and the resulting conclusions are discussed.

This report is intended as an overview of post-closure performance assessment methods for high-level radioactive waste repositories and is designed to give the reader a broad sense of the state of the art of this technology. As described here, ''the state of the art'' includes only what has been reported in report, journal, and conference proceedings literature through August 1987. There is a very large literature on the performance of high-levelwaste repositories. In order to make a review of this breadth manageable, its scope must be carefully defined. The essential principle followed is that only methods of calculating the long-term performance of waste repositories are described. The report is organized to reflect, in a generalized way, the logical order to steps that would be taken in a typical performance assessment. Chapter 2 describes ways of identifying scenarios and estimating their probabilities. Chapter 3 presents models used to determine the physical and chemical environment of a repository, including models of heat transfer, radiation, geochemistry, rock mechanics, brine migration, radiation effects on chemistry, and coupled processes. The next two chapters address the performance of specific barriers to release of radioactivity. Chapter 4 treats engineered barriers, including containers, waste forms, backfills around waste packages, shaft and borehole seals, and repository design features. Chapter 5 discusses natural barriers, including ground water systems and stability of salt formations. The final chapters address optics of general applicability to performance assessment models. Methods of sensitivity and uncertainty analysis are described in Chapter 6, and natural analogues of repositories are treated in Chapter 7. 473 refs., 19 figs., 2 tabs.

Six alloys are being considered as possible materials for the fabrication of containers for the disposal of high-level radioactive waste. Three of these candidate materials are copper-based alloys: CDA 102 (oxygen-free copper), CDA 613 (Cu-7Al), and CDA 715 (Cu-30Ni). The other three are iron- to nickel-based austenitic materials: Types 304L and 316L stainless steels and Alloy 825. Radioactive waste will include spent-fuel assemblies from reactors as well as waste in borosilicate glass and will be sent to the prospective site at Yucca Mountain, Nevada, for disposal. The waste-package containers must maintain substantially complete containment for at least 300 yr and perhaps as long as 1000 yr. During the first 50 yr after emplacement, the containers must be retrievable from the disposal site. Shortly after emplacement of the containers in the repository, they will be exposed to high temperatures and high gamma radiation fields from the decay of high-levelwaste. This radiation will promote the radiolytic decomposition of moist air to hydrogen. This volume surveys the available data on the effects of hydrogen on the six candidate alloys for fabrication of the containers. For copper, the mechanism of hydrogen embrittlement is discussed, and the effects of hydrogen on the mechanical properties of the copper-based alloys are reviewed. The solubilities and diffusivities of hydrogen are documented for these alloys. For the austenitic materials, the degradation of mechanical properties by hydrogen is documented. The diffusivity and solubility of hydrogen in these alloys are also presented. For the copper-based alloys, the ranking according to resistance to detrimental effects of hydrogen is: CDA 715 (best) > CDA 613 > CDA 102 (worst). For the austenitic alloys, the ranking is: Type 316L stainless steel {approx} Alloy 825 > Type 304L stainless steel (worst). 87 refs., 19 figs., 8 tabs.

The 324 Building in the Hanford 300 Area contains Radiochemical Engineering Cells and HighLevel Vault tanks (the {open_quotes}REC/HLV{close_quotes}) for research and development activities involving radioactive materials. Radioactive mixed waste within this research installation, found primarily in B-Cell and three of the highlevel vault tanks, is subject to RCRA/DWR ({open_quotes}RCRA{close_quotes}) regulations for storage. This white paper provides a baseline RCRA compliance summary of MW management in the REC/HLV, based on best available knowledge. The REC/HLV compliance project, of which this paper is a part, is intended to achieve the highest degree of compliance practicable given the special technical difficulties of managing high activity radioactive materials, and to assure protection of human health and safety and the environment. The REC/HLV was constructed in 1965 to strict standards for the safe management of highly radioactive materials. Mixed waste in the REC/HLV consists of discarded tools and equipment, dried feed stock from nuclear waste melting experiments, contaminated particulate matter, and liquid feed stock from various experimental programs in the vault tanks. B-Cell contains most of these materials. Total radiological inventory in B-Cell is estimated at 3 MCi, about half of which is potentially {open_quotes}dispersible{close_quotes}, that is, it is in small pieces or mobile particles. Most of the mixed waste currently in the REC/HLV was generated or introduced before mixed wastes were subjected to RCRA in 1987.

This paper describes the results of a collaborative technology development program, sponsored by the Tanks Focus Area, to use electrochemical noise (EN) for corrosion monitoring in underground storage tanks. These tanks, made of carbon or stainless steels, contain high-level radioactive liquid waste (HLW) generated by weapons production or radioactive liquid waste from nuclear fuel reprocessing activities at several Department of Energy (DOE) sites. The term EN is used to describe low frequency fluctuations in current and voltage measurements associated with corrosion. In their most basic form, EN-based corrosion monitoring systems measure and record these fluctuations over time from electrodes immersed in the environment of interest--in this case, radioactive tank waste. The resulting EN signals have characteristic patterns for different corrosion mechanisms. In recent years, engineers and scientists from several DOE sites, in collaboration with several private companies, have conducted laboratory studies and field applications to correlate the EN signals with corrosion mechanisms active in the radioactive waste tanks. The participating DOE sites are Hanford, Savannah River, Oak Ridge Reservation and the Idaho National Engineering and Environmental Laboratory. The commercial vendors have included HiLine Engineering and Fabrication, Inc., EIC Laboratories, Inc., and AEA Technologies. Successful deployment of the EN technology will yield improved information of waste tank corrosion conditions, better tank management, and lower overall cost.

This paper describes the results of a collaborative technology development program, sponsored by the Tanks Focus Area, to use electrochemical noise (EN) for corrosion monitoring in underground storage tanks. These tanks, made of carbon or stainless steels, contain high-level radioactive liquid waste (HLW) generated by weapons production or radioactive liquid waste from nuclear fuel reprocessing activities at several Department of Energy (DOE) sites. The term EN is used to describe low frequency fluctuations in current and voltage measurements associated with corrosion. In their most basic form, EN-based corrosion monitoring systems measure and record these fluctuations over time from electrodes immersed in the environment of interest - in this case, radioactive tank waste. The resulting EN signals have characteristic patterns for different corrosion mechanisms. In recent years, engineers and scientists from several DOE sites, in collaboration with several private companies, have conducted laboratory studies and field applications to correlate the EN signals with corrosion mechanisms active in the radioactive waste tanks. The participating DOE sites are Hanford, Savannah River, Oak Ridge Reservation and the Idaho National Engineering and Environmental Laboratory. The commercial vendors have included HiLine Engineering and Fabrication, Inc., EIC Laboratories, Inc., and M A Technologies. Successful deployment of the EN technology will yield improved information of waste tank corrosion conditions, better tank management, and lower overall cost.

The removal of cesium radioisotope (137Cs) from the HighLevelWaste stored in underground storage tanks at the Hanford site is a formidable chemical separations challenge for the Waste Treatment Plant. An eluatable organic-based ion exchange resin was selected as the baseline technology (1). The baseline technology design employs a proprietary macrocyclic weak-acid ion exchange resin to adsorb the cesium (137Cs) during the process loading cycle in a fixed bed column design. Following loading, the cesium is eluted from the resin using a nitric acid eluant. Previous work provided limited understanding of the performance of the resin, processed with actual wastes, and under multiple load and elute conditions, which are required for the ion exchange technology to be underpinned sufficiently for resolution of all process-related design issues before flowsheet and construction drawings can be released. By performing multiple ion exchange column tests with waste feeds, and measuring the chemical and radionuclide compositions of the waste feeds, column effluents and column eluants, ion exchange stream composition information can be provided for supporting resolution of selected design issues.

The objective of this work is to develop a more detailed mechanistic understanding of the thermal reactions that lead to gas production in certain high-levelwaste storage tanks at the Hanford, Washington site. Prediction of the combustion hazard for these wastes and engineering parameters for waste processing depend upon both a knowledge of the composition of stored wastes and the changes that they undergo as a result of thermal and radiolytic decomposition. Since 1980 when Delagard first demonstrated that gas production (H2and N2O initially, later N2 and NH3)in the affected tanks was related to oxidative degradation of metal complexants present in the waste, periodic attempts have been made to develop detailed mechanisms by which the gases were formed. These studies have resulted in the postulation of a series of reactions that account for many of the observed products, but which involve several reactions for which there is limited, or no, precedent. For example, Al(OH)4 has been postulated to function as a Lewis acid to catalyze the reaction of nitrite ion with the metal complexants, NO is proposed as an intermediate, and the ratios of gaseous products may be a result of the partitioning of NO between two or more reactions. These reactions and intermediates have been the focus of this project since its inception in 1996.

The West Valley Demonstration Project (WVDP) is nearing completion of radioactive high-levelwaste (HLW) retrieval from its storage tanks and subsequent vitrification of the HLW into borosilicate glass. Currently, 99.5% of the sludge radioactivity has been recovered from the storage tanks and vitrified. Waste recovery of cesium-137 (Cs-137) adsorbed on a zeolite media during waste pretreatment has resulted in 97% of this radioactivity being vitrified. Approximately 84% of the original 1.1 x 1018 becquerels (30 million curies) of radioactivity was efficiently vitrified from July 1996 to June 1998 during Phase I processing. The recovery of the last 16% of the waste has been challenging due to a number of factors, primarily the complex internal structural support system within the main 2.8 million liter (750,000 gallon) HLW tank designated 8D-2. Recovery of this last waste has become exponentially more challenging as less and less HLW is available to mobilize and transfer to the Vitrification Facility. This paper describes the progressively more complex techniques being utilized to remove the final small percentage of radioactivity from the HLW tanks, and the multiple characterization technologies deployed to determine the quantity of Cs-137, strontium-90 (Sr-90), and alpha-transuranic (alpha-TRU) radioactivity remaining in the tanks.

Laboratory tests have established the feasibility of using neutron activation of copper as a means for measuring the moisture in Hanford`s high-level radioactive waste tanks. The performance of the neutron activation technique to measure moisture is equivalent to the neutron moisture gauges or neutron logs commonly used in commercial well-logging. The principle difference is that the activation of {sup 64}Cu (t{sub 1/2} = 12.7 h) replaces the neutron counters used in moisture gauges or neutron logs. For application to highly radioactive waste tanks, the Cu activation technique has the advantage that it is insensitive to very strong gamma radiation fields or high temperatures. In addition, this technique can be deployed through tortuous paths or in confined spaces such as within the bore of a cone penetrometer. However, the results are not available in ``real-time``. The copper probe`s sensitivity to moisture was measured using simulated tank waste of known moisture content. This report describes the preparation of the simulated waste mixtures and the experiments performed to demonstrate the capabilities of the neutron activation technique. These experiments included determination of the calibration curve of count rate versus moisture content using a single copper probe, measurement of the calibration curve based on ``near-field `` to ``far-field`` counting ratios using a multiple probe technique, and profiling the activity of the copper probe as a function of the vertical height within a simulated waste barrel.

The Department of Energy`s program for disposing of nuclear High-LevelWaste (HLW) and transuranic (TRU) waste has been impeded by overwhelming political opposition fueled by public perceptions of actual risk. Analysis of these perceptions shows them to be deeply rooted in images of fear and dread that have been present since the discovery of radioactivity. The development and use of nuclear weapons linked these images to reality and the mishandling of radioactive waste from the nations military weapons facilities has contributed toward creating a state of distrust that cannot be erased quickly or easily. In addition, the analysis indicates that even the highly educated technical community is not well informed on the latest technology involved with nuclear HLW and TRU waste disposal. It is not surprising then, that the general public feels uncomfortable with DOE`s management plans for with nuclear HLW and TRU waste disposal. Postponing the permanent geologic repository and use of Monitored Retrievable Storage (MRS) would provide the time necessary for difficult social and political issues to be resolved. It would also allow time for the public to become better educated if DOE chooses to become proactive.

Stainless steel-zirconium (SS-Zr) alloys have been developed for the consolidation and disposal of waste stainless steel, zirconium, and noble metal fission products such as Nb, Mo, Tc, Ru, Pd, and Ag recovered from spent nuclear fuel assemblies. These remnant waste metals are left behind following electrometallurgical treatment, a molten salt-based process being demonstrated by Argonne National Laboratory. Two SS-Zr compositions have been selected as baseline waste form alloys: (a) stainless steel-15 wt% zirconium (SS-15Zr) for stainless steel-clad fuels and (b) zirconium-8 wt% stainless steel (Zr-8SS) for Zircaloy-clad fuels. Simulated waste form alloys were prepared and tested to characterize the metallurgy of SS-15Zr and Zr-8SS and to evaluate their physical properties and corrosion resistance. Both SS-15Zr and Zr-8SS have multi-phase microstructures, are mechanically strong, and have thermophysical properties comparable to other metals. They also exhibit high resistance to corrosion in simulated groundwater as determined by immersion, electrochemical, and vapor hydration tests. Taken together, the microstructure, physical property, and corrosion resistance data indicate that SS-15Zr and Zr-8SS are viable materials as high-levelwaste forms.

Technical aspects related to the development of standards for regulating geologic disposal of highlevel radioactive waste, with particular emphasis on the sources and magnitudes of uncertainties associated with methods for predicting post closure repository performance and potential health risks to future generations are discussed. Important licensing and regulatory issues involved in geologic waste disposal, and the social and political climate in which issues of highlevelwaste management are being debated are also considered.

This paper describes the results of the analyses of HighLevelWaste (HLW) sludge slurry samples and of the calculations necessary to decay the radionuclides to meet the reporting requirement in the Waste Acceptance Product Specifications (WAPS) [1]. The concentrations of 45 radionuclides were measured. The results of these analyses provide input for radioactive decay calculations used to project the radionuclide inventory at the specified index years, 2015 and 3115. This information is necessary to complete the Production Records at Savannah River Site's Defense Waste Processing Facility (DWPF) so that the final glass product resulting from Macrobatch 5 (MB5) can eventually be submitted to a Federal Repository. Five of the necessary input radionuclides for the decay calculations could not be measured directly due to their low concentrations and/or analytical interferences. These isotopes are Nb-93m, Pd-107, Cd-113m, Cs-135, and Cm-248. Methods for calculating these species from concentrations of appropriate other radionuclides will be discussed. Also the average age of the MB5 HLW had to be calculated from decay of Sr-90 in order to predict the initial concentration of Nb-93m. As a result of the measurements and calculations, thirty-one WAPS reportable radioactive isotopes were identified for MB5. The total activity of MB5 sludge solids will decrease from 1.6E+04 {micro}Ci (1 {micro}Ci = 3.7E+04 Bq) per gram of total solids in 2008 to 2.3E+01 {micro}Ci per gram of total solids in 3115, a decrease of approximately 700 fold. Finally, evidence will be given for the low observed concentrations of the radionuclides Tc-99, I-129, and Sm-151 in the HLW sludges. These radionuclides were reduced in the MB5 sludge slurry to a fraction of their expected production levels due to SRS processing conditions.

Cold war plutonium production led to extensive amounts of radioactive waste stored in tanks at the Department of Energy's (DOE) Hanford site. Bechtel National, Inc. is building the largest nuclear Waste Treatment Plant in the world located at the Department of Energy's Hanford site to immobilize the millions of gallons of radioactive waste. The site comprises five main facilities; Pretreatment, HighLevelWaste vitrification, Low Active Waste vitrification, an Analytical Lab and the Balance of Facilities. The pretreatment facilities will separate the high and low level waste. The highlevelwaste will then proceed to the HLW facility for vitrification. Vitrification is a process of utilizing a melter to mix molten glass with radioactive waste to form a stable product for storage. The melter cave is designated as the HighLevelWaste Melter Cave Support Handling System (HSH). There are several key processes that occur in the HSH cell that are necessary for vitrification and include: feed preparation, mixing, pouring, cooling and all maintenance and repair of the process equipment. Due to the cell's highlevel radiation, remote handling equipment provided by PaR Systems, Inc. is required to install and remove all equipment in the HSH cell. The remote handling crane is composed of a bridge and trolley. The trolley supports a telescoping tube set that rigidly deploys a TR 4350 manipulator arm with seven degrees of freedom. A rotating, extending, and retracting slewing hoist is mounted to the bottom of the trolley and is centered about the telescoping tube set. Both the manipulator and slewer are unique to this cell. The slewer can reach into corners and the manipulator's cross pivoting wrist provides better operational dexterity and camera viewing angles at the end of the arm. Since the crane functions will be operated remotely, the entire cell and crane have been modeled with 3-D software. Model simulations have been used to confirm operational and maintenance

Cracks of various shapes and sizes exist in large high-levelwaste (HLW) tanks at several DOE sites. There is justifiable concern that these cracks could grow to become unstable causing a substantial release of liquid contaminants to the environment. Accurate prediction of crack growth behavior in the tanks, especially during accident scenarios, is not possible with existing analysis methodologies. This research project responds to this problem by developing an improved ability to predict crack growth in material structure combinations that are ductile (Fig. 1). This new model not only addresses the problem for these tanks, but also has applicability to any crack in any ductile structure.

The Subseabed Disposal Project (SDP) was part of an international program that investigated the feasibility of high-level radioactive waste disposal in the deep ocean sediments. This report briefly describes the seven-step iterative performance assessment procedures used in this study and presents representative results of the last iteration. The results of the performance are compared to interim standards developed for the SDP, to other conceptual repositories, and to related metrics. The attributes, limitations, uncertainties, and remaining tasks in the SDP feasibility phase are discussed.

A series of cyclic potentiodynamic polarization tests was performed on samples of ASTM A537 carbon steel in support of a probability-based approach to evaluate the effect of chloride and sulfate on corrosion the steel's susceptibility to pitting corrosion. Testing solutions were chosen to systemically evaluate the influence of the secondary aggressive species, chloride, and sulfate, in the nitrate based, high-levelwastes. The results suggest that evaluating the combined effect of all aggressive species, nitrate, chloride, and sulfate, provides a consistent response for determining corrosion susceptibility. The results of this work emphasize the importance for not only nitrate concentration limits, but also chloride and sulfate concentration limits.

A key component of the Engineered Barrier System (EBS) being designed for containment of spent-fuel and high-levelwaste at the proposed geological repository at Yucca Mountain, Nevada is a two-layer canister. In this particular design, the inner barrier is made of a corrosion resistant material (CRM) such as Alloy 625 or C-22, while the outer barrier is made of a corrosion-allowance material (CAM) such as carbon steel or Alloy 400. An integrated predictive model is being developed to account for the effects of localized environmental conditions in the CRM-CAM crevice on the initiation and propagation of pits through the CRM.

Washington River Protection Solutions (WRPS), operator of HighLevel Radioactive Waste (HLW) Tank Farms at the Hanford Site, is taking an over 20-year leap in technology, replacing systems that were monitored with clipboards and obsolete computer systems, as well as solving major operations and maintenance hurdles in the area of process automation and information management. While WRPS is fully compliant with procedures and regulations, the current systems are not integrated and do not share data efficiently, hampering how information is obtained and managed.

Large volumes of high-levelwaste (HLW) currently stored in tanks at DOE sites contain both sludges and supernatants. The sludges are composed of insoluble precipitates of actinides, radioactive fission products, and nonradioactive components. The supernatants are alkaline carbonate solutions, which can contain soluble actinides, fission products, metal ions, and high concentrations of major electrolytes including sodium hydroxide, nitrate, nitrite, phosphate, carbonate, aluminate, sulfate, and organic complexants. The organic complexants include several compounds that can form strong aqueous complexes with actinide species and fission products including ethylenediaminetetraacetic acid (EDTA), N-(2-hydroxyethyl)ethylenediaminetriacetic acid (HEDTA), nitrilotriacetic acid (NTA), iminodiacetic acid (IDA), citrate, glycolate, gluconate, and degradation products, formate and oxalate.

Large volumes of high-levelwaste (HLW) currently stored in tanks at DOE sites contain both sludges and supernatants. The sludges are composed of insoluble precipitates of actinides, radioactive fission products, and nonradioactive components. The supernatants are alkaline carbonate solutions, which can contain soluble actinides, fission products, metal ions, and high concentrations of major electrolytes including sodium hydroxide, nitrate, nitrite, phosphate, carbonate, aluminate, sulfate, and organic complexants. The organic complexants include several compounds that can form strong aqueous complexes with actinide species and fission products including ethylenediaminetetraacetic acid (EDTA), N-(2-hydroxyethyl)ethylenediaminetriacetic acid (HEDTA), nitrilotriacetic acid (NTA), iminodiacetic acid (IDA), citrate, glycolate, gluconate, and degradation products, formate and oxalate.

Dispersion modeling of non-radiological airborne emissions associated with the construction and operation of three alternatives for high-levelwaste salt disposition at the Savannah River Site has been completed. The results will be used by Department of Energy-Savannah River in the preparation of the salt disposition supplemental environmental impact statement. Estimated maximum ground-level concentrations of applicable regulated air pollutants of the site boundary and at the distance to a hypothetical, co-located onsite worker are summarized in tables. In all cases, model estimated ambient concentrations are less than regulatory standards.

The concentrated medium and high-level liquid radio chemicals effluents contain nitric acid, water along with the dissolved chemicals including the nitrates of the radio nuclides. Highlevel liquid waste contain mainly nitrates of cesium, strontium, cerium, zirconium, chromium, barium, calcium, cobalt, copper, pickle, iron etc. and other fission products. This concentrated solution requires further evaporation, dehydration, drying and decomposition in temperature range of 150 to 700 deg. C. The addition of the calcined solids in vitrification pot, instead of liquid feed, helps to avoid low temperature zone because the vaporization of the liquid and decomposition of nitrates do not take place inside the melter. In our work Differential and thermo gravimetric studies has been carried out in the various stages of thermal treatment including drying, dehydration and conversion to oxide forms. Experimental studies were done to characterize the chemicals present in high-level radioactive waste. A Rotary Ball Kiln Calciner was used for development of the process because this is amenable for continuous operation and moderately good heat transfer can be achieved inside the kiln. This also has minimum secondary waste and off gases generation. The Rotary Ball Kiln Calciner Demonstration facility system was designed and installed for the demonstration of calcination process. The Rotary Ball Kiln Calciner is a slowly rotating slightly inclined horizontal tube that is externally heated by means of electric resistance heating. The liquid feed is sprayed onto the moving bed of metal balls in a slowly rotating calciner by a peristaltic type-metering pump. The vaporization of the liquid occurs in the pre-calcination zone due to counter current flow of hot gases. The dehydration and denitration of the solids occurs in the calcination zone, which is externally heated by electrical furnace. The calcined powder is cooled in the post calcination portion. It has been demonstrated that the

This report documents the initial conceptual design studies for a repository sealing system for a high-level nuclear waste repository in salt. The first step in the initial design studies was to review the current design level, termed schematic designs. This review identified practicality of construction and development of a design methodology as two key issues for the conceptual design. These two issues were then investigated during the initial design studies for seal system materials, seal placement, backfill emplacement, and a testing and monitoring plan. The results of these studies have been used to develop a program plan for completion of the sealing system conceptual design. 60 refs., 26 figs., 18 tabs.

Spent nuclear fuel and high-level radioactive waste are materials from nuclear power plants and government defense programs. These materials contain highly radioactive elements, such as cesium, strontium, technetium, and neptunium. Some of these elements will remain radioactive for a few years, while others will be radioactive for millions of years. Exposure to such radioactive materials can cause human health problems. Scientists worldwide agree that the safest way to manage these materials is to dispose of them deep underground in what is called a geologic repository.

PUREX high-levelwaste (HLW) is contained at the West Valley Demonstration Project (WVDP) in an underground carbon-steel storage tank. The HLW consists of a precipitated sludge and an alkaline supernate. This report describes the system that the WVDP has developed and implemented to resuspend and wash the HLW sludge from the tank. The report discusses Sludge Mobilization and Wash System (SMWS) equipment design, installation, and testing. The storage tank required modifications to accommodate the SMWS. These modifications are discussed as well.

The present plan for disposal of high-level radioactive waste in the US is to seal it in containers before emplacement in a geologic repository. A proposed site at Yucca Mountain, Nevada, is being evaluated for its suitability as a geologic repository. The containers will probably be made of either an austenitic or a copper-based alloy. Models of alloy degradation are being used to predict the long-term performance of the containers under repository conditions. The models are of uniform oxidation and corrosion, localized corrosion, and stress corrosion cracking, and are applicable to worst-case scenarios of container degradation. This paper reviews several of the models.

These proceedings are arranged into six broad categories: general overview of analysis and design; characterization of faulting; characterization of design ground vibratory ground motion; considerations for underground facilities; considerations for surface facilities; and guidelines for instrumentation and monitoring. Discussions are given on the relative merits and inadequacies of state-of-the-art design/analysis practices and methodologies in the seismic and dynamic analysis and design field in relation to high-level nuclear waste repositories. All papers have been processed for inclusion on the data base.

A multiphase experimental program to assess and understand waste glass behavior under a wide range of conditions has been in progress at the Savannah River Laboratory (SRL) for over a decade. An important part of this experimental effort is to assess the effects of repository relevant conditions on performance of SRSwaste glass, in both controlled laboratory tests, as well as in actual field experiments. In laboratory test, SRSwaste glass, simulated and in many cases also fully radioactive, has been tested in environments of salt, basalt, shale, granite, clay and tuff. In field experiments, there are four joint international programs being conducted in four different countries, involving burial of SRS simulated waste glass in granite, limestone, clay and salt geologies. This report discusses the SRSwaste glass studies in limestone at Ballidon, UK..

During the development of the feed processing flowsheet for the Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS), research had shown that use of formic acid (HCOOH) could accomplish several processing objectives with one chemical addition. These objectives included the decomposition of tetraphenylborate, chemical reduction of mercury, production of acceptable rheological properties in the feed slurry, and controlling the oxidation state of the glass melt pool. However, the DEPF research had not shown that some vitrification slurry feeds had a tendency to evolve hydrogen (H{sub 2}) and ammonia (NH{sub 3}) as the result of catalytic decomposition of CHOOH with noble metals (rhodium, ruthenium, palladium) in the feed. Testing conducted at Pacific Northwest Laboratory and later at the Savannah River Technical Center showed that the H{sub 2} and NH{sub 3} could evolve at appreciable rates and quantities. The explosive nature of H{sub 2} and NH{sub 3} (as ammonium nitrate) warranted significant mitigation control and redesign of both facilities. At the time the explosive gas evolution was discovered, the DWPF was already under construction and an immediate hardware fix in tandem with flowsheet changes was necessary. However, the Hanford Waste Vitrification Plant (HWVP) was in the design phase and could afford to take time to investigate flowsheet manipulations that could solve the problem, rather than a hardware fix. Thus, the HWVP began to investigate alternatives to using HCOOH in the vitrification process. This document describes the selection, evaluation criteria, and strategy used to evaluate the performance of the alternative chemical additives to CHOOH. The status of the evaluation is also discussed.

Reliable instrumentation will be needed to evaluate the characteristics of proposed high-level nuclear-wasted-repository sites and to monitor the performance of selected sites during the operational period and into repository closure. A study has been done to assess the reliability of instruments used in Department of Energy (DOE) waste repository related experiments and in other similar geological applications. The study included experiences with geotechnical, hydrological, geochemical, environmental, and radiological instrumentation and associated data acquisition equipment. Though this paper includes some findings on the reliability of instruments in each of these categories, the emphasis is on experiences with geotechnical instrumentation in hostile repository-type environments. We review the failure modes, rates, and mechanisms, along with manufacturers modifications and design changes to enhance and improve instrument performance; and include recommendations on areas where further improvements are needed.

The President signed the Congressional Joint Resolution on July 23, 2002, that designated the Yucca Mountain site for a proposed geologic repository to dispose of the nation's spent nuclear fuel (SNF) and high-level radioactive waste (HLW). The United States (U.S.) Department of Energy's (DOE) Office of Civilian Radioactive Waste Management (OCRWM) is currently focusing its efforts on submitting a license application to the U.S. Nuclear Regulatory Commission (NRC) in December 2004 for construction of the proposed repository. The legislative framework underpinning the U.S. repository program is the basis for its continuity and success. The repository development program has significantly benefited from international collaborations with other nations in the Americas.

The cleanup of highlevel defense nuclear waste at the Hanford site presents several progressive challenges. Among these is the removal and disposal of various components from buried active waste tanks to allow new equipment insertion or hazards mitigation. A unique automated retrieval system at the tank provides for retrieval, high pressure washing, inventory measurement, and containment for disposal. Key to the inventory measurement is a three detector HPGe high performance gamma spectroscopy system capable of recovering data at up to ninety per cent saturation (200,000 counts per second). Data recovery is based on a unique embedded electronic pulser and specialized software to report the inventory. Each of the detectors have different shielding specified through Monte Carlo simulation with the MCNP program. This shielding provides performance over a dynamic range of eight orders of magnitude. System description, calibration issues and operational experiences are discussed.

This paper describes an approach to identifying source terms from possible accidents during the preclosure phase of a high-level nuclear waste repository. A review of the literature on repository safety analyses indicated that source term estimation is in a preliminary stage, largely based on judgement-based scoping analyses. The approach developed here was to partition the accident space into domains defined by certain threshold values of temperature and impact energy density which may arise in potential accidents and specify release fractions of various radionuclides, present in the waste form, in each domain. Along with a more quantitative understanding of accident phenomenology, this approach should help in achieving a clearer perspective on scenarios important to preclosure safety assessments of geologic repositories. 18 refs., 3 tabs.

Magnetic measurements constitute a promising method for the characterization of nuclear waste glasses in view of their simplicity and small sample weight requirements. Initial studies of simulated high-levelwaste glasses show that the Curie constant is generally a useful indicator of the Fe{sup 2+}:Fe{sup 3+} ratio. Glasses produced by air-cooling in large vessels show systematic deviations between experimental and calcined values, which are indicative of the presence of small amounts of crystalline iron-containing phases. Most of the iron in these phases becomes dissolved in the glass upon re-heating and more rapid quenching. The studies further show that upon leaching the glass in water some of the iron in the surface regions of the glass is converted to a form which has high temperature-independent magnetic susceptibility.

The convective transfer of radionuclides by subsurface water from a geological repository of solidified high-level radioactive wastes (HLW) is considered. The repository is a cluster of wells of large diameter with HLW disposed of in the lower portions of the wells. The safe distance between wells as a function of rock properties and parameters of well loading with wastes has been estimated from mathematical modeling. A maximum permissible concentration of radionuclides in subsurface water near the ground surface above the repository is regarded as a necessary condition of safety. The estimates obtained show that well repositories allow for a higher density of solid HLW disposal than shaft storage facilities. Advantages and disadvantages of both types of storage facilities are considered in order to estimate the prospects for their use for underground disposal of solid HLW.

The proposed national high-level nuclear waste repository will be designed to store approximately 70,000 tons of commercial spent fuel, but other forms of waste will also be considered for ultimate storage at this site. Plutonium in the form of PuO{sub 2} may be added to borosilicate glass for ultimate disposal in the repository. The maximum amount of this fissile that may be added to a glass ``log`` will be limited by its ability to sustain a chain reaction. In this study, the removal of neutron absorbers from a glass log and the subsequent possibility of water infiltration were studied to find corresponding neutron multiplication factors. Weight fractions of 1%, 2%, and 3% PuO{sub 2} were analyzed in the study. The results show the maximum amount of plutonium fissile that may be safely added to a glass log under conditions that lead to leaching of the principal neutron absorbers from the glass.

The Site Recommendation (SR) process for the potential repository for spent nuclear fuel (SNF) and high-level nuclear waste (HLW) at Yucca Mountain, Nevada is underway. Fulfillment of the requirements for substantially complete containment of the radioactive waste emplaced in the potential repository and subsequent slow release of radionuclides from the Engineered Barrier System (EBS) into the geosphere will rely on a robust waste container design, among other EBS components. Part of the SR process involves sensitivity studies aimed at elucidating which model parameters contribute most to the drip shield and waste package degradation characteristics. The model parameters identified included (a) general corrosion rate model parameters (temperature-dependence and uncertainty treatment), and (b) stress corrosion cracking (SCC) model parameters (uncertainty treatment of stress and stress intensity factor profiles in the Alloy 22 waste package outer barrier closure weld regions, the SCC initiation stress threshold, and the fraction of manufacturing flaws oriented favorably for through-wall penetration by SCC). These model parameters were reevaluated and new distributions were generated. Also, early waste package failures due to improper heat treatment were added to the waste package degradation model. The results of these investigations indicate that the waste package failure profiles are governed by the manufacturing flaw orientation model parameters and models used.

The Defense Waste Processing Facility processes legacy nuclear waste generated at the Savannah River Site during production of enriched uranium and plutonium required by the Cold War. The nuclear waste is first treated via a complex sequence of controlled chemical reactions and then vitrified into a borosilicate glass form and poured into stainless steel canisters. Converting the nuclear waste into borosilicate glass is a safe, effective way to reduce the volume of the waste and stabilize the radionuclides. One of the constituents in the nuclear waste is mercury, which is present because it served as a catalyst in the dissolution of uranium-aluminum alloy fuel rods. At high temperatures mercury is corrosive to off-gas equipment, this poses a major challenge to the overall vitrification process in separating mercury from the waste stream prior to feeding the high temperature melter. Mercury is currently removed during the chemical process via formic acid reduction followed by steam stripping, which allows elemental mercury to be evaporated with the water vapor generated during boiling. The vapors are then condensed and sent to a hold tank where mercury coalesces and is recovered in the tank's sump via gravity settling. Next, mercury is transferred from the tank sump to a purification cell where it is washed with water and nitric acid and removed from the facility. Throughout the chemical processing cell, compounds of mercury exist in the sludge, condensate, and off-gas; all of which present unique challenges. Mercury removal from sludge waste being fed to the DWPF melter is required to avoid exhausting it to the environment or any negative impacts to the Melter Off-Gas system. The mercury concentration must be reduced to a level of 0.8 wt% or less before being introduced to the melter. Even though this is being successfully accomplished, the material balances accounting for incoming and collected mercury are not equal. In addition, mercury has not been effectively

The Department of Energy (DOE) has specific technical and documentation requirements for high-levelwaste (HLW) that is to be placed in a federal repository. This document describes in general terms the strategy to be used at the Idaho National Engineering and Environmental Laboratory (INEEL) to demonstrate that vitrified HLW, if produced at the INEEL, meets these requirements. Waste form, canister, quality assurance, and documentation specifications are discussed. Compliance strategy is given, followed by an overview of how this strategy would be implemented for each specification.

In this paper, the structure and properties of the different solid forms currently being developed for high-level radioactive waste disposal are compared. Good capacity to accept all the elements in the waste and flexibility of composition range to accommodate variations in the waste, are primarily discussed. 13 refs.

High-levelwaste from spent commercial PWR fuel was vitrified during the Nuclear Waste Vitrification Project (NWVP), conducted at Pacific Northwest Laboratory in 1978 and 1979. The solidification equipment employed sintered metal filters and a conventional wet off-gas cleanup system to decontaminate the gases generated during waste processing. This report is a summary of a study completed under contract with U.S. Department of Energy (DOE) to determine the fate of volatile fission products in the filters and off-gas system. The study objectives were to determine (1) the extent of volatile fission product deposition within the sintered metal filters, (2) the distribution of fission products in the off-gas system, and (3) the factors that control the distribution. Such information is needed for the design and operation of future waste immobilization systems. The scope of the study included identification, preparation, and microprobe analyses of suitable filter samples and radiochemical analyses of off-gas system samples. The report describes the methods of sample selection and gathering, the analytical methods and results, and the interpretations of the results. Relationships to vitrification system operations are defined and operating conditions to minimize problems recommended. 37 figures, 14 tables.

The remediation of the 1 {times} 10{sup 8} gal of highly radioactive waste in the underground storage tanks (USTs) at five US Department of Energy (DOE) sites is one of DOE`s greatest challenges. Therefore, the DOE Office of Environmental Management has created the Tank Focus Area (TFA) to manage an integrated technology development program that results in the safe and efficient remediation of UST waste. The TFA has divided its efforts into five areas, which are safety, characterization, retrieval/closure, pretreatment, and immobilization. All DOE pretreatment activities are integrated by the Pretreatment Technical Integration Manager of the TFA. For FY 1996, the 14 pretreatment tasks are divided into 3 systems: supernate separations, sludge treatment, and solid/liquid separation. The plans and recent results of these TFA tasks, which include two 25,000-gal demonstrations and two former TFA tasks on Cs removal, are presented. The pretreatment goals are to minimize the volume of high-levelwaste and the radioactivity in low-level waste.

ABSTRACT A new monitoring system was developed at Pacific Northwest National Laboratory (PNNL) to quickly generate real-time data/analysis to facilitate a timely response to the dynamic characteristics of a radioactive highlevelwaste stream. The developed process monitor features Raman and Coriolis/conductivity instrumentation configured for the remote monitoring, MatLab-based chemometric data processing, and comprehensive software for data acquisition/storage/archiving/display. The monitoring system is capable of simultaneously and continuously quantifying the levels of all the chemically significant anions within the waste stream including nitrate, nitrite, phosphate, carbonate, chromate, hydroxide, sulfate, and aluminate. The total sodium ion concentration was also determined independently by modeling inputs from on-line conductivity and density meters. In addition to the chemical information, this monitoring system provides immediate real-time data on the flow parameters, such as flow rate and temperature, and cumulative mass/volume of the retrieved waste stream. The components and analytical tools of the new process monitor can be tailored for a variety of complex mixtures in chemically harsh environments, such as pulp and paper processing liquids, electroplating solutions, and radioactive tank wastes. The developed monitoring system was tested for acceptability before it was deployed for use in Hanford Tank S-109 retrieval activities. The acceptance tests included performance inspection of hardware, software, and chemometric data analysis to determine the expected measurement accuracy for the different chemical species that are encountered during S-109 retrieval.

A new monitoring system was developed at Pacific Northwest National Laboratory (PNNL) to quickly generate real-time data/analysis to facilitate a timely response to the dynamic characteristics of a radioactive highlevelwaste stream. The developed process monitor features Raman and Coriolis/conductivity instrumentation configured for the remote monitoring, MatLab-based chemometric data processing, and comprehensive software for data acquisition/storage/archiving/display. The monitoring system is capable of simultaneously and continuously quantifying the levels of all the chemically significant anions within the waste stream including nitrate, nitrite, phosphate, carbonate, chromate, hydroxide, sulfate, and aluminate. The total sodium ion concentration was also determined independently by modeling inputs from on-line conductivity and density meters. In addition to the chemical information, this monitoring system provides immediate real-time data on the flow parameters, such as flow rate and temperature, and cumulative mass/volume of the retrieved waste stream. The components and analytical tools of the new process monitor can be tailored for a variety of complex mixtures in chemically harsh environments, such as pulp and paper processing liquids, electroplating solutions, and radioactive tank wastes. The developed monitoring system was tested for acceptability before it was deployed for use in Hanford Tank S-109 retrieval activities. The acceptance tests included performance inspection of hardware, software, and chemometric data analysis to determine the expected measurement accuracy for the different chemical species that are encountered during S-109 retrieval. (authors)

Since 1963, calcination of aqueous wastes from reprocessing of DOE-owned spent nuclear fuels has resulted in the accumulation of approximately 3800 m{sup 3} of high-levelwaste (HLW) at the Idaho Chemical Processing Plant (ICPP). The waste is in the form of a granular solid called calcine and is stored on site in stainless steel bins which are encased in concrete. Due to the leachability of {sup 137}Cs and {sup 90}Sr and possibly other radioactive components, the calcine is not suitable for final disposal. Hence, a process to immobilize calcine in glass is being developed. Since radioactive components represent less than 1 wt % of the calcine, separation of actinides and fission products from inert components is being considered to reduce the volume of HLW requiring final disposal. Current estimates indicate that compared to direct vitrification, a volume reduction factor of 10 could result in significant cost savings. Aqueous processes, which involve calcine dissolution in nitric acid followed by separation of actinide and fission products by solvent extraction and ion exchange methods, are being developed. Pyrochemical separation methods, which generate small volumes of aqueous wastes and do not require calcine dissolution, have been evaluated as alternatives to aqueous processes. This report describes three proposed pyrochemical flowsheets and presents the results of experimental studies conducted to evaluate their feasibility. The information presented is a consolidation of three reports, which should be consulted for experimental details.

Glycolic acid is being studied as a replacement for formic acid in the Defense Waste Processing Facility (DWPF) feed preparation process. After implementation, the recycle stream from DWPF back to the high-levelwaste Tank Farm will contain soluble sodium glycolate. Most of the potential impacts of glycolate in the Tank Farm were addressed via a literature review and simulant testing, but several outstanding issues remained. This report documents the actual-waste tests to determine the impacts of glycolate on storage and evaporation of Savannah River Site high-levelwaste. The objectives of this study are to address the following: Determine the extent to which sludge constituents (Pu, U, Fe, etc.) dissolve (the solubility of sludge constituents) in the glycolate-containing 2H-evaporator feed. Determine the impact of glycolate on the sorption of fissile (Pu, U, etc.) components onto sodium aluminosilicate solids. The first objective was accomplished through actual-waste testing using Tank 43H and 38H supernatant and Tank 51H sludge at Tank Farm storage conditions. The second objective was accomplished by contacting actual 2H-evaporator scale with the products from the testing for the first objective. There is no anticipated impact of up to 10 g/L of glycolate in DWPF recycle to the Tank Farm on tank waste component solubilities as investigated in this test. Most components were not influenced by glycolate during solubility tests, including major components such as aluminum, sodium, and most salt anions. There was potentially a slight increase in soluble iron with added glycolate, but the soluble iron concentration remained so low (on the order of 10 mg/L) as to not impact the iron to fissile ratio in sludge. Uranium and plutonium appear to have been supersaturated in 2H-evaporator feed solution mixture used for this testing. As a result, there was a reduction of soluble uranium and plutonium as a function of time. The change in soluble uranium concentration was

The extraction behavior of actinides like U(VI), Pu(IV), and Am(III) as well as of fission products like Tc(VII), Zr(IV), Eu(III), and structural material Fe(III) using nitric acid/simulated pressurized heavy water reactor-high-levelwaste solution as aqueous phase and N,N'-dimethyl-N,N'dibutyl tetra decyl malonamide (DMDBTDMA) in n-dodecane as solvent was investigated. The present work contributes significantly toward the development of nonphosphorous (environmentally friendly) extractant capable of partitioning long-lived actinides and fission products like {sup 99}Tc from relatively short lived fission products like {sup 90}Sr and {sup 137}Cs as well as inactive components present in high-levelwaste. The D{sub Am} values determined in the temperature range between 15 and 45 deg. C showed a gradual decrease with increase in temperature throughout the acidity range. The effect of N,N di-2-ethylhexyl acetamide (D2EHAA) as phase modifier on the physicochemical properties of DMDBTDMA/n-dodecane solvent was investigated and could be correlated with parameters like the limiting organic concentration value of HNO{sub 3}, D{sub Am}, or phase disengagement time. Overall comparison of diamide extraction (Diamex) and transuranium extraction (Truex) solvents has been made.

This report summarizes the bottoms-up cost estimates for fabrication of high-level radioactive waste disposal containers based on the Site Characterization Plan Conceptual Design (SCP-CD). These estimates were acquired by Babcock and Wilcox (B&S) under sub-contract to Lawrence Livermore National Laboratory (LLNL) for the Yucca Mountain Site Characterization Project (YMP). The estimates were obtained for two leading container candidate materials (Alloy 825 and CDA 715), and from other three vendors who were selected from a list of twenty solicited. Three types of container designs were analyzed that represent containers for spent fuel, and for vitrified high-levelwaste (HLW). The container internal structures were assumed to be AISI-304 stainless steel in all cases, with an annual production rate of 750 containers. Subjective techniques were used for estimating QA/QC costs based on vendor experience and the specifications derived for the LLNL-YMP Quality Assurance program. In addition, an independent QA/QC analysis is reported which was prepared by Kasier Engineering. Based on the cost estimates developed, LLNL recommends that values of $825K and $62K be used for the 1991 TSLCC for the spent fuel and HLW containers, respectively. These numbers represent the most conservative among the three vendors, and are for the high-nickel anstenitic steel (Alloy 825). 6 refs., 7 figs.

Over a decade ago, an in-tank precipitation process to remove Cs-137 from radioactive highlevelwaste (HLW) supernates was demonstrated at the Savannah River Site (SRS). The full scale demonstration with actual HLW was performed in SRS Tank 48 (T48). Sodium tetraphenylborate (NaTPB) was added to enable Cs-137 extraction as CsTPB. The CsTPB, an organic, and its decomposition products proved to be problematic for subsequent processing of the Cs-137 precipitate in the SRS HLW vitrification facility for ultimate disposal in a HLW repository. Fluidized Bed Steam Reforming (FBSR) is being considered as a technology for destroying the organics and nitrates in the T48 waste to render it compatible with subsequent HLW vitrification. During FBSR processing the T48 waste is converted into organic-free and nitrate-free carbonate-based minerals which are water soluble. The soluble nature of the carbonate-based minerals allows them to be dissolved and pumped to the vitrification facility or returned to the tank farm for future vitrification. The initial use of the FBSR process for T48 waste was demonstrated with simulated waste in 2003 at the Savannah River National Laboratory (SRNL) using a specially designed sealed crucible test that reproduces the FBSR pyrolysis reactions, i.e. carbonate formation, organic and nitrate destruction. This was followed by pilot scale testing of simulants at the Science Applications International Corporation (SAIC) Science & Technology Application Research (STAR) Center in Idaho Falls, ID by Idaho National Laboratory (INL) and SRNL in 2003-4 and then engineering scale demonstrations by THOR{reg_sign} Treatment Technologies (TTT) and SRS/SRNL at the Hazen Research, Inc. (HRI) test facility in Golden, CO in 2006 and 2008. Radioactive sealed crucible testing with real T48 waste was performed at SRNL in 2008, and radioactive Benchscale Steam Reformer (BSR) testing was performed in the SRNL Shielded Cell Facility (SCF) in 2008.

This report presents a preconceptual design study for processing radioactive high-level liquid waste presently stored in underground tanks at Western New York Nuclear Service Center (WNYNSC) near West Valley, New York, and for incorporating the radionculides in that waste into a solid. The high-level liquid waste accumulated from the operation of a chemical reprocessing plant by the Nuclear Fuel Services, Inc. from 1966 to 1972. The high-level liquid waste consists of approximately 560,000 gallons of alkaline waste from Purex process operations and 12,000 gallons of acidic (nitric acid) waste from one campaign of processing thoria fuels by a modified Thorex process (during this campaign thorium was left in the waste). The alkaline waste contains approximately 30 million curies and the acidic waste contains approximately 2.5 million curies. The reference process described in this report is concerned only with chemically processing the high-level liquid waste to remove radionuclides from the alkaline supernate and converting the radionuclide-containing nonsalt components in the waste into a borosilicate glass.

A high-levelwaste (HLW) monitor has been designed, fabricated and tested at full-scale for deployment inside a Hanford tank. The Solid-Liquid Interface Monitor (SLIM) integrates a commercial sonar system with a mechanical deployment system for deploying into an underground waste tank. The system has undergone several design modifications based upon changing requirements at Hanford. We will present the various designs of the monitor from first to last and will present performance data from the various prototype systems. We will also present modeling of stresses in the enclosure under 85 mph wind loading. The system must be able to function at winds up to 15 mph and must withstand a maximum loading of 85 mph. There will be several examples presented of engineering tradeoffs made as FIU analyzed new requirements and modified the design to accommodate. We will present our current plans for installing into the Cold Test Facility at Hanford and into a double-shelled tank at Hanford. Finally, we will present our vision for how this technology can be used at Hanford and Savannah River Site to improve the filling and emptying of high-levelwaste tanks. In conclusion: 1. The manually operated first-generation SLIM is a viable option on tanks where personnel are allowed to work on top of the tank. 2. The remote controlled second-generation SLIM can be utilized on tanks where personnel access is limited. 3. The totally enclosed fourth-generation SLIM, when the design is finalized, can be used when the possibility exists for wind dispersion of any HLW that maybe on the system. 4. The profiling sonar can be used effectively for real-time monitoring of the solid-liquid interface over a large area. (authors)

Three iron- to nickel-based austenitic alloys (Types 304L and 316L stainless steels and Alloy 825) are being considered as candidate materials for the fabrication of high-level radioactive-waste containers. Waste will include fuel assemblies from reactors as well as high-levelwaste in borosilicate glass forms, and will be sent to the prospective repository at Yucca Mountain, Nevada. The decay of radionuclides in the repository will result in the generation of substantial heat and in fluences of gamma radiation. Container materials may undergo any of several modes of degradation in this environment, including atmospheric oxidation; uniform aqueous phase corrosion; pitting; crevice corrosion; sensitization and intergranular stress corrosion cracking (IGSCC); and transgranular stress corrosion cracking (TGSCC). This report is an analysis of data relevant to the pitting, crevice corrosion, and stress corrosion cracking (SCC) of the three austenitic candidate alloys. The candidates are compared in terms of their susceptibilities to these forms of corrosion. Although all three candidates have demonstrated pitting and crevice corrosion in chloride-containing environments, Alloy 825 has the greatest resistance to these types of localized corrosion (LC); such resistance is important because pits can penetrate the metal and serve as crack initiation sites. Both Types 304L and 316L stainless steels are susceptible to SCC in acidic chloride media. In contrast, SCC has not been documented in Alloy 825 under comparable conditions. Gamma radiation has been found to enhance SCC in Types 304 and 304L stainless steels, but it has no detectable effect on the resistance of Alloy 825 to SCC. Furthermore, while the effects of microbiologically induced corrosion have been observed for 300-series stainless steels, nickel-based alloys such as Alloy 825 seem to be immune to such problems. 211 refs., 49 figs., 10 tabs.

The Hanford radioactive tank waste will be separated into low-activity waste and high-levelwaste that will both be vitrified into borosilicate glasses. To demonstrate the feasibility of vitrification and the durability of the high-levelwaste glass, a high-levelwaste sample from Tank AZ-101 was processed to glass in a hot cell and analyzed with respect to chemical composition, radionuclide content, waste loading, and the presence of crystalline phases and then tested for leachability. The glass was analyzed with inductively coupled plasma-atomic emission spectroscopy, inductively coupled plasma-mass spectrometry, γ energy spectrometry, α spectrometry, and liquid scintillation counting. The WISE Uranium Project calculator was used to calculate the main sources of radioactivity to the year 3115. The observed crystallinity and the results of leachability testing of the glass will be reported in Part 2 of this paper.

The Hanford radioactive tank waste will be separated into low-activity waste and high-levelwaste that will both be vitrified into borosilicate glasses. To demonstrate the feasibility of vitrification and the durability of the high-levelwaste glass, a high-levelwaste sample from Tank AZ-101 was processed to glass, analyzed with respect to chemical composition, radionuclide content, waste loading, and the presence of crystalline phases and then tested for leachability. The glass was analyzed with inductively coupled plasma-atomic emission spectroscopy, inductively coupled plasma-mass spectrometry, γ-energy spectrometry, α-spectrometry, and liquid scintillation counting. The WISE Uranium Project calculator was used to calculate the main sources of radioactivity to the year 3115. The observed crystallinity and the results of leachability testing of the glass will be reported in Part 2 of this paper.

This report evaluates the feasibility of disposing U.S. high-level radioactive waste in granite several hundred meters below the surface of the earth. The U.S. has many granite formations with positive attributes for permanent disposal. Similar crystalline formations have been extensively studied by international programs, two of which, in Sweden and Finland, are the host rocks of submitted or imminent repository license applications. This report is enabled by the advanced work of the international community to establish functional and operational requirements for disposal of a range of waste forms in granite media. In this report we develop scoping performance analyses, based on the applicable features, events, and processes (FEPs) identified by international investigators, to support generic conclusions regarding post-closure safety. Unlike the safety analyses for disposal in salt, shale/clay, or deep boreholes, the safety analysis for a mined granite repository depends largely on waste package preservation. In crystalline rock, waste packages are preserved by the high mechanical stability of the excavations, the diffusive barrier of the buffer, and favorable chemical conditions. The buffer is preserved by low groundwater fluxes, favorable chemical conditions, backfill, and the rigid confines of the host rock. An added advantage of a mined granite repository is that waste packages would be fairly easy to retrieve, should retrievability be an important objective. The results of the safety analyses performed in this study are consistent with the results of comprehensive safety assessments performed for sites in Sweden, Finland, and Canada. They indicate that a granite repository would satisfy established safety criteria and suggest that a small number of FEPs would largely control the release and transport of radionuclides. In the event the U.S. decides to pursue a potential repository in granite, a detailed evaluation of these FEPs would be needed to inform site

Attempts to solve the problem of high-levelwaste disposal including the spent fuel from nuclear power plants have been made in the Czech Republic for over the 10 years. Already in 1991 the Ministry of Environment entitled The Czech Geological Survey to deal with the siting of the locality for HLW disposal and the project No. 3308 ''The geological research of the safe disposal of highlevelwaste'' had started. Within this project a sub-project ''A selection of perspective HLW disposal sites in the Bohemian Massif'' has been elaborated and 27 prospective areas were identified in the Czech Republic. This selection has been later narrowed to 8 areas which are recently studied in more detail. As a parallel research activity with siting a granitic body Melechov Massif in Central Moldanubian Pluton has been chosen as a test site and the 1st stage of research i.e. evaluation and study of its geological, hydrogeological, geophysical, tectonic and structural properties has been already completed. The Melechov Massif was selected as a test site after the recommendation of WATRP (Waste Management Assessment and Technical Review Programme) mission of IAEA (1993) because it represents an area analogous with the host geological environment for the future HLW and spent fuel disposal in the Czech Republic, i.e. variscan granitoids. It is necessary to say that this site would not be in a locality where the deep repository will be built, although it is a site suitable for oriented research for the sampling and collection of descriptive data using up to date and advanced scientific methods. The Czech Republic HLW and spent fuel disposal programme is now based on The Concept of Radioactive Waste and Spent Nuclear Fuel Management (''Concept'' hereinafter) which has been prepared in compliance with energy policy approved by Government Decree No. 50 of 12th January 2000 and approved by the Government in May 2002. Preparation of the Concept was required, amongst other reasons in

Consideration of the technical feasibility of Yucca Mountain in Nevada as the site for a high-level nuclear waste repository has led to an intense debate regarding the economic, social, and political impacts of the repository. Impediments to the siting process mean that the nuclear waste problem is being resolved by adhering to the status quo, in which nuclear waste is stored at scattered sites near major population centers. To assess the merits of alternative siting strategies--including both the permanent repository and the status quo- we consider the variables that would be included in a model designed to select (1) the optimal number of disposal facilities, (2) the types of facilities (e.g., permanent repository or monitored retrievable facility), and (3) the geographic location of storage sites. The objective function in the model is an all-inclusive measure of social cost. The intent of the exercise is not to demonstrate the superiority of any single disposal strategy; uncertainties preclude a conclusive proof of optimality for any of the disposal options. Instead, we want to assess the sensitivity of a variety of proposed solutions to variations in the physical, economic, political, and social variables that influence a siting strategy.

AREVA has more than 30 years experience in operating industrial HLW (HighLevel radioactive Waste) vitrification facilities (AVM - Marcoule Vitrification Facility, R7 and T7 facilities). This vitrification technology was based on borosilicate glasses and induction-heating. AVM was the world's first industrial HLW vitrification facility to operate in-line with a reprocessing plant. The glass formulation was adapted to commercial Light Water Reactor fission products solutions, including alkaline liquid waste concentrates as well as platinoid-rich clarification fines. The R7 and T7 facilities were designed on the basis of the industrial experience acquired in the AVM facility. The AVM vitrification process was implemented at a larger scale in order to operate the R7 and T7 facilities in-line with the UP2 and UP3 reprocessing plants. After more than 30 years of operation, outstanding record of operation has been established by the R7 and T7 facilities. The industrial startup of the CCIM (Cold Crucible Induction Melter) technology with enhanced glass formulation was possible thanks to the close cooperation between CEA and AREVA. CCIM is a water-cooled induction melter in which the glass frit and the waste are melted by direct high frequency induction. This technology allows the handling of highly corrosive solutions and high operating temperatures which permits new glass compositions and a higher glass production capacity. The CCIM technology has been implemented successfully at La Hague plant.

The Cementitious Barriers Partnership (CBP) is focused on reducinguncertainties in current methodologies for assessing cementitious barrier performanceand increasing the consistency and transparency in the assessment process. Oneimportant set of US Department of Energy challenges is assessing the integrity andclosure of the high-levelwaste (HLW) tanks that currently store millions of gallons ofhighly radioactive wastes. Many of these tanks are decades past their design lives, haveleaked or been overfilled, and must be emptied and closed to satisfy regulatoryagreements. Carbonation-induced corrosion has been identified as a primary degradationand possible failure mechanism for the HLW tanks prior to closure. After closure theimpact of carbonation (and concurrent oxidation) may be to increase the release andshort-range transport of contaminants of concern. HLW tanks may be significantlyempty for many years (and possibly decades) prior to closure; the performance of theclosed tank over centuries, if not millennia, must be assessed to evaluate the potentialrelease of residual radionuclides to the environment. CBP is developing models to evaluate a representative HLW tank closure scenarioincluding the potential impacts of carbonation on waste tanks prior to and post closure.CBP modeling tools, including LeachXS™/ORCHESTRA, are being used to simulatewaste tank carbonation, major constituent leaching, and contaminant releases to evaluatethe source term and near-field conditions. Simulations presented here include sensitivityanalysis for uncracked concrete to varying input parameters including composition,effective diffusivities, and thermodynamic parameters.

This document updates the planning basis for the U.S. Department of Energy (DOE) to provide a sufficient quantity of high-levelwaste (HLW) feed to the privatization contractor during Phase I. The DOE has sufficient HLW feed to satisfy the minimum order quantity and, with the transfer system upgrades recommended by Galbraith (1996), the means to provide the feed to the private contractor`s facility during Phase I. Assuming the planned DST retrieval system (two 300-hp mixer pumps), the total sludge available as feed for Waste Envelope D is at least 130 percent of the minimum order quantity specified in the RFP, assuming dilute caustic washing. However, additional sources of feed will need to be identified to support the optional extension of HLW processing in Phase I. The maximum HLW sludge inventory available in the three DSTs identified for Phase I is less 1514 than the maximum order quantity (460 MT versus 465 MT). Assuming the anticipated efficiencies of the DST retrieval systems are achieved, the amount of waste available for feed would only be 70 to 80 percent of the maximum order quantity.

The Integrated DWPF Melter System (IDMS), operated by the Savannah River Laboratory, is a pilot scale facility used in support of the start-up and operation of the Department of Energy`s Defense Waste Processing Facility. The IDMS has successfully demonstrated, on an engineering scale (one-fifth), that simulated highlevel radioactive waste (HLW) sludge can be chemically treated with formic acid to adjust both its chemical and physical properties, and then blended with simulated precipitate hydrolysis aqueous (PHA) product and borosilicate glass frit to produce a melter feed which can be processed into a durable glass product. The simulated sludge, PHA and frit were blended, based on a product composition program, to optimize the loading of the waste glass as well as to minimize those components which can cause melter processing and/or glass durability problems. During all the IDMS demonstrations completed thus far, the melter feed and the resulting glass that has been produced met all the required specifications, which is very encouraging to future DWPF operations. The IDMS operations also demonstrated that the volatile components of the melter feed (e.g., mercury, nitrogen and carbon, and, to a lesser extent, chlorine, fluorine and sulfur) did not adversely affect the melter performance or the glass product.

This report evaluates the feasibility of high-level radioactive waste disposal in shale within the United States. The U.S. has many possible clay/shale/argillite basins with positive attributes for permanent disposal. Similar geologic formations have been extensively studied by international programs with largely positive results, over significant ranges of the most important material characteristics including permeability, rheology, and sorptive potential. This report is enabled by the advanced work of the international community to establish functional and operational requirements for disposal of a range of waste forms in shale media. We develop scoping performance analyses, based on the applicable features, events, and processes identified by international investigators, to support a generic conclusion regarding post-closure safety. Requisite assumptions for these analyses include waste characteristics, disposal concepts, and important properties of the geologic formation. We then apply lessons learned from Sandia experience on the Waste Isolation Pilot Project and the Yucca Mountain Project to develop a disposal strategy should a shale repository be considered as an alternative disposal pathway in the U.S. Disposal of high-level radioactive waste in suitable shale formations is attractive because the material is essentially impermeable and self-sealing, conditions are chemically reducing, and sorption tends to prevent radionuclide transport. Vertically and laterally extensive shale and clay formations exist in multiple locations in the contiguous 48 states. Thermal-hydrologic-mechanical calculations indicate that temperatures near emplaced waste packages can be maintained below boiling and will decay to within a few degrees of the ambient temperature within a few decades (or longer depending on the waste form). Construction effects, ventilation, and the thermal pulse will lead to clay dehydration and deformation, confined to an excavation disturbed zone within

Liquidus temperature (TL) studies of high-Zr high-levelwaste (HLW) borosilicate glasses have identified three primary phases: baddelyite (ZrO2), zircon (ZrSiO4), and alkali-zirconium silicates, such as parakeldyshite (Na2ZrSi2O7). Using published TL data for HLW glasses with these primary phases, we have computed partial specific TLs for major glass components. On the Na2O-SiO2-ZrO2 submixture, we have determined approximate positions of the boundaries between the baddelyite, zircon, and parakeldyshite primary phase fields. The maximum that can dissolve at 1150?C in a borosilicate HLW glass subjected to common processability and acceptability constraints appears to be 16.5 mass% ZrO2.

Tectonic characterization of a potential high-level nuclear waste repository at Yucca Mountain, Nevada, is needed to assess seismic and possible volcanic hazards that could affect the site during the preclosure (next 100 years) and the behavior of the hydrologic system during the postclosure (the following 10,000 years) periods. Tectonic characterization is based on assembling mapped geological structures in their chronological order of development and activity, and interpreting their dynamic interrelationships. Addition of mechanistic models and kinematic explanations for the identified tectonic processes provides one or more tectonic models having predictive power. Proper evaluation and application of tectonic models can aid in seismic design and help anticipate probable occurrence of future geologic events of significance to the repository and its design.

A key component of the Engineered Barrier System (EBS) being designed for containment of spent-fuel and high-levelwaste at the proposed geological repository at Yucca Mountain, Nevada is a two-layer canister. In this particular design, the inner barrier is made of a corrosion resistant material (CRM) such as Alloy 825, 625 or C-22, while the outer barrier is made of a corrosion-allowance material (CAM) such as A516 or Monel 400. At the present time, Alloy C-22 and A516 are favored. This publication addresses the development of models to account for corrosion of Alloy C-22 surfaces exposed directly to the Near Field Environment (NFE), as well as to the exacerbated conditions in the CAM-CRM crevice.

Reprocessing of irradiated fuel assemblies generates canisters filled with vitrified highlevel radioactive waste (HLW). The canisters are made of stainless steel and welded leak-tight. These HLW canisters are subject to transport from the fuel reprocessing plant to intermediate and final storage. Since 1983, a number of different packages based on the type B(U) concept of the IAEA were therefore designed, tested, licensed and manufactured for the transport and the interim storage of HLW canisters in flasks. The theoretical layout of the cask was tested by measurements on a prototype cask of the scale 1:1. The measured heat transfer characteristics of this newly developed transport and storage cask will be described. They can be written as a function Nu - C * Ra{sup m}, with the variable factor C, and the constant exponent m., The factor C is different for the vertical and the horizontal position of the cask.

This paper describes a safety analysis of a transfer process for high-level radioactive and toxic waste. The analysis began with a hazard assessment that used elements of What If, Checklist, Failure Modes and Effects Analysis, and Hazards and Operability Study (HAZOP) techniques to identify and rough-in accident sequences. Based on this preliminary analysis, the most significant accident sequences were developed further using event trees. Quantitative frequency estimates for the accident sequences were based on operational data taken from the historical record of the site where the process is performed. Several modeling challenges were encountered in the course of the study. These included linked initiating and accident progression events, fire propagation modeling, accounting for administrative control violations, and handling mission-phase effects.

Nickel based alloys are candidate materials for the storage of highlevelwaste (HLW) generated from reprocessing of spent nuclear fuel. In the present investigation Alloy 690 and Alloy 693 are assessed by potentiodynamic anodic polarization technique for their corrosion behavior in 3 M HNO 3, 3 M HNO 3 containing simulated HLW and in chloride medium. Both the alloys were found to possess good corrosion resistance in both the media at ambient condition. Microstructural examination was carried out by SEM for both the alloys after electrolytic etching. Compositional analysis of the passive film formed on the alloys in 3 M HNO 3 and 3 M HNO 3 with HLW was carried out by XPS. The surface of Alloy 690 and Alloy 693, both consists of a thin layer of oxide of Ni, Cr, and Fe under passivation in both the media. The results of investigation are presented in the paper.

Types 304L, 316L, and 321 austenitic stainless steel and Incoloy 825 are being considered as candidate container materials for emplacing high-levelwaste in a tuff repository. The stress corrosion cracking susceptibility of these materials under simulated tuff repository conditions was evaluated by using the notched C-ring method. The tests were conducted in boiling synthetic groundwater as well as in the steam/air phase above the boiling solutions. All specimens were in contact with crushed Topopah Spring tuff. The investigation showed that microcracks are frequently observed after testing as a result of stress corrosion cracking or intergranular attack. Results showing changes in water chemistry during test are also presented.

This paper provides a regulatory perspective from the viewpoint of the potential licensee, the U.S. Department of Energy (DOE), on the National Academy of Sciences (NAS) report on Yucca Mountain standards issued in August 1995, and on how the recommendations in that report should be considered in the development of high-level radioactive waste standards applicable to Yucca Mountain. The paper first provides an overview of the DOE perspective and then discusses several of the issues that are of most importance in the development of the regulatory framework for Yucca Mountain, including both the U.S. Environmental Protection Agency (EPA) standard and the U.S. Nuclear Regulatory Commission (NRC) implementing regulation. These issues include: the regulatory time frame, the risk/dose limit, the definition of the reference biosphere, human intrusion, and natural processes and events.

Evaluation of potential high-level nuclear waste repository sites is an area where geophysical capabilities and limitations may significantly impact a major governmental program. Since there is concern that extensive exploratory drilling might degrade most potential disposal sites, geophysical methods become crucial as the only nondestructive means to examine large volumes of rock in three dimensions. Characterization of potential sites requires geophysicists to alter their usual mode of thinking: no longer are anomalies being sought, as in mineral exploration, but rather their absence. Thus the size of features that might go undetected by a particular method take on new significance. Legal and regulatory considerations that stem from this different outlook, most notably the requirements of quality assurance (necessary for any data used in support of a repository license application), are forcing changes in the manner in which geophysicists collect and document their data. -Authors

Two primary challenges to characterizing Hanford’s high-level radioactive waste slurry prior to transfer to a treatment facility are the ability to representatively sample million-gallon tanks and to estimate the critical velocity of the complex slurry. Washington River Protection Solutions has successfully demonstrated a sampling concept that minimizes sample errors by collecting multiple sample increments from a sample loop where the mixed tank contents are recirculated. Pacific Northwest National Laboratory has developed and demonstrated an ultrasonic-based Pulse-Echo detection device that is capable of detecting a stationary settled bed of solids in a pipe with flowing slurry. These two concepts are essential elements of a feed delivery strategy that drives the Hanford clean-up mission.

This report describes the results of a test conducted by Battelle to assess the effects of inhibited water washing and caustic leaching on the composition of the Hanford tank C-106 high-levelwaste (HLW) solids. The objective of this work was to determine the composition of the C-106 solids remaining after washing with 0.01M NaOH or leaching with 3M NaOH. Another objective of this test was to determine the solubility of various C-106 components as a function of temperature. The work was conducted according to test plan BNFL-TP-29953-8,Rev. 0, Determination of the Solubility of HLW Sludge Solids. The test went according to plan, with only minor deviations from the test plan. The deviations from the test plan are discussed in the experimental section.

The Glass Material Oxidation and Dissolution System (GMODS) is a recently invented process for the direct, single-step conversion of spent nuclear fuel (SNF) to high-levelwaste (HLW) glass. GMODS converts metals, ceramics, organics, and amorphous solids to glass in a single step. Conventional vitrification technology can not accept feeds containing metals or carbon. The GMODS has the potential to solve several issues associated with the disposal of various US Department of Energy (DOE) miscellaneous SNFs: (1) chemical forms unacceptable for repository disposal; (2) high cost of qualifying small quantities of particular SNFs for disposal; (3) limitations imposed by high-enriched SNF in a repository because of criticality and safeguards issues; and (4) classified design information. Conversion of such SNFs to glass eliminates these concerns. A description of the GMODS, {open_quotes}strawman{close_quotes} product criteria, experimental work to date, and product characteristics are included herein.

The US Department of Energy (DOE) has millions of gallons of highlevel nuclear waste stored in underground tanks at Hanford, Washington and Savannah River, South Carolina. These tanks will eventually be emptied and decommissioned. This will leave a residue of sludge adhering to the interior tank surfaces that may contaminate groundwaters with radionuclides and RCRA metals. Experimentation on such sludges is both dangerous and prohibitively expensive so there is a great advantage to developing artificial sludges. The US DOE Environmental Management Science Program (EMSP) has funded a program to investigate the feasibility of developing such materials. The following text reports on the success of this program, and suggests that much of the radioisotope inventory left in a tank will not move out into the surrounding environment. Ultimately, such studies may play a significant role in developing safe and cost effective tank closure strategies.

The age-related degradation mechanisms which affect the concrete and the reinforcing steel in the high-levelwaste (HLW) storage tanks art evaluated with respect to their potential significance to the continued performance of the concrete, and am classified into non-significant and potentially significant. The identified potentially significant degradation mechanisms include the effects of elevated temperature, freezing and thawing, leaching of calcium hydroxide, aggressive chemical attack, and corrosion of the reinforcing steel. To the extent that available knowledge permits, these mechanisms are generically evaluated and quantified so that site-specific plans may be developed to verify whether significant degradation has occurred in the concrete, and, if so, to formulate mitigating measures to avoid further deterioration and possibly repair the degradation or pursue other management options.

This volume surveys the effects of welding on the degradation modes of three austenitic alloys: Types 304L and 316L stainless steels and Alloy 825. These materials are candidates for the fabrication of containers for the long-term storage of high-level nuclear waste. The metallurgical characteristics of fusion welds are reviewed here and related to potential degradation modes of the containers. Three specific areas are discussed in depth: (1) decreased resistance to corrosion in the forms of preferential corrosion, sensitization, and susceptibility to stress corrosion cracking, (2) hot cracking in the heat-affected zone and the weld zone, and (3) formation of intermetallic phases. The austenitic alloys are ranked as follows in terms of overall weldability: Alloy 825 (best) > Type 316L stainless steel > Type 304L stainless steel (worst). 108 refs., 31 figs., 7 tabs.

The Seabed Disposal Project (SDP) was evaluating the technical feasibility of high-level nuclear waste disposal in deep ocean sediments. Working standards were needed for risk assessments, evaluation of alternative designs, sensitivity studies, and conceptual design guidelines. This report completes a three part program to develop radiological standards for the feasibility phase of the SDP. The characteristics of subseabed disposal and how they affect the selection of standards are discussed. General radiological protection standards are reviewed, along with some new methods, and a systematic approach to developing standards is presented. The selected interim radiological standards for the SDP and the reasons for their selection are given. These standards have no legal or regulatory status and will be replaced or modified by regulatory agencies if subseabed disposal is implemented. 56 refs., 29 figs., 15 tabs.

Tectonic characterization of a potential high-level nuclear waste repository at Yucca Mountain, Nevada, is needed to assess seismic and possible volcanic hazards that could affect the site during the preclosure (next 100 years) and the behavior of the hydrologic system during the postclosure (the following 10,000 years) periods. Tectonic characterization is based on assembling mapped geological structures in their chronological order of development and activity, and interpreting their dynamic interrelationships. Addition of mechanistic models and kinematic explanations for the identified tectonic processes provides one or more tectonic models having predictive power. Proper evaluation and application of tectonic models can aid in seismic design and help anticipate probable occurrence of future geologic events of significance to the repository and its design.

Spinel crystals precipitate in high-levelwaste glasses containing Fe, Cr, Ni , Mn, Zn, and Ru. The liquidus temperature (TL) of spinel as the primary crystallization phase is a function of glass composition and the spinel solubility (c0) is a function of both glass composition and temperature (T). Previously reported models of TL as a function of composition are based on TL measured directly, which requires laborious experimental procedures. Viewing the curve of c0 versus T as the liquidus line allows a significant broadening of the composition region for model fitting. This paper estimates TL as a function of composition based on c0 data obtained with the X-ray diffraction technique.

This discussion focuses on the use of borehole to borehole geophysical measurements to detect geological discontinuities in HighLevelWaste (HLW) repository sites. The need for these techniques arises from: (a) the requirement that a HLW repository's characteristics and projected performance be known with a high degree of confidence; and (b) the inadequacy of other geophysical methods in mapping fractures. Probing configurations which can be used to characterize HLW sites are described. Results from experiments in which these techniques were applied to problems similar to those expected at repository sites are briefly discussed. The use of a procedure designed to reduce uncertainty associated with all geophysical exploration techniques is proposed; key components of the procedure are defined.

The interaction of aluminophosphate glass with water at 95°C for 35 days results in glass heterogenization and in the appearance of a gel layer and various phases. The leaching rate of elements is low owing to the formation of a protective layer on the glass surface. It is shown that over 80% of uranium leached from the glass matrix occurs as colloids below 450 nm in size characterized by high migration ability in the geological environment. To determine the composition of these colloids is a primary task for further studies. Water vapor is a crystallization factor for glasses. The conditions as such may appear even at early stages of glass storage because of the failure of seals on containers of high-level radioactive wastes. The examination of water resistance of crystallized matrices and determination of the fraction of radionuclide in colloids are also subjects for further studies.

Bechtel BWXT Idaho, Maintenance and Operating Contractor for the Department of Energy at the Idaho National Engineering and Environmental Laboratory, has emptied, cleaned, and sampled six of the eleven 1.135 x 10(6) L highlevelwaste underground storage tanks at the Idaho Nuclear Technology and Engineering Center, well ahead of the State of Idaho Consent Order cleaning schedule. Cleaning of a seventh tank is expected to be complete by the end of calendar year 2004. The tanks, with associated vaults, valve boxes, and distribution systems, are being closed to meet Resource Conservation and Recovery Act regulations and Department of Energy orders. The use of remotely operated equipment placed in the tanks through existing tank riser access points, sampling methods and application of as-low-as-reasonably-achievable (ALARA) principles have proven effective in keeping personnel dose low during equipment removal, tank, vault, and valve box cleaning, and sampling activities, currently at 0.03 Sv. PMID:15824589

Spinel crystals precipitate in high-levelwaste glasses containing Fe, Cr, Ni, Mn, Zn, and Ru. The liquidus temperature (T{sub L}d) of spinel as the primary crystallization phase is a function of glass composition, and the spinel solubility (c{sub o}) is a function of both glass composition and temperature (T). Previously reported models of T{sub L} as a function of composition are based on T{sub L} measured directly, which requires laborious experimental procedures. Viewing the curve of c{sub o} versus T as the liquidus line allows a significant broadening of the composition region for model fitting. This paper estimates T{sub L} as a function of composition based on c{sub o} data obtained with the X-ray diffraction technique.

A key component of the Engineered Barrier System (EBS) being designed for containment of spent-fuel and high-levelwaste at the proposed geological respository at Yucca Mountain, Nevada is a two-layer canister. In this particular design, the inner barrier is made of a corrosion resistant material (CRM) such as Alloy 825, 625 or C-22, while the outer barrier is made of a corrosion-allowance material (CAM) such as A516 or Monel 400. At the present time, Alloy C-22 and A516 are favored. This publication addresses the development of models to account for corrosion of Alloy C-22 surfaces exposed directly to the Near Field Environmental (NFE), as well as to the exacerbated conditions in the CAM-CRM crevice.

Guidelines are provided for developing programs to promote the structural integrity of high-levelwaste storage tanks and transfer lines at the facilities of the Department of Energy. Elements of the program plan include a leak-detection system, definition of appropriate loads, collection of data for possible material and geometric changes, assessment of the tank structure, and non-destructive examination. Possible aging degradation mechanisms are explored for both steel and concrete components of the tanks, and evaluated to screen out nonsignificant aging mechanisms and to indicate methods of controlling the significant aging mechanisms. Specific guidelines for assessing structural adequacy will be provided in companion documents. Site-specific structural integrity programs can be developed drawing on the relevant portions of the material in this document.

The primary objective of the task summarized in this paper was to demonstrate a methodology for evaluating alternative strategies for preclosure lay-up of the two high-levelwaste (HLW) storage tanks at the West Valley Demonstration Project (WVDP). Lay-up is defined as the period between operational use of tanks for waste storage and final closure. The U.S. Department of Energy (DOE) is planning to separate the environmental impact statement (EIS) for completion of closure of the WVDP into two separate EISs. The first EIS will cover only waste management and decontamination. DOE expects to complete this EIS in about 18 months. The second EIS will cover final decommissioning and closure and may take up to five years to complete. This approach has been proposed to expedite continued management of the waste and decontamination activities in advance of the final EIS and its associated Record of Decision on final site closure. Final closure of the WVDP site may take 10 to 15 years; therefore, the tanks need to be placed in a safe, stable condition with minimum surveillance during an extended lay-up period. The methodology developed for ranking the potential strategies for lay-up of the WVDP tanks can be used to provide a basis for a decision on the preferred path forward. The methodology is also applicable to determining preferred lay-up approaches at other DOE sites. Some of the alternative strategies identified for the WVDP should also be considered for implementation at the other DOE sites. Each site has unique characteristics that would require unique considerations for lay-up.

'Flammable gases can be generated in DOE high-levelwaste tanks, including radiolytic hydrogen, and during cesium precipitation from salt solutions, benzene. Under normal operating conditions the potential for deflagration or detonation from these gases is precluded by purging and ventilation systems, which remove the flammable gases and maintain a well-mixed condition in the tanks. Upon failure of the ventilation system, due to seismic or other events, however, it has proven more difficult to make strong arguments for well-mixed conditions, due to the potential for density-induced stratification which can potentially sequester fuel or oxidizer at concentrations significantly higher than average. This has complicated the task of defining the safety basis for tank operation. Waste-tank mixing processes have considerable overlap with similar large-enclosure mixing processes that occur in enclosure fires and nuclear reactor containments. Significant differences also exist, so that modeling techniques that have been developed previously can not be directly applied to waste tanks. In particular, mixing of air introduced through tank roof penetrations by buoyancy and pressure driven exchange flows, mixed convection induced by an injected high-velocity purge jet interacting with buoyancy driven flow, and onset and breakdown of stable stratification under the influence of an injected jet have not been adequately studied but are important in assessing the potential for accumulation of high-concentration pockets of fuel and oxygen. Treating these phenomena requires a combination of experiments and the development of new, more general computational models than those that have been developed for enclosure fires. U.C. Berkeley is now completing the second year of its three-year project that started in September, 1996. Excellent progress has been made in several important areas related to waste-tank ventilation and mixing processes.'

A program to resolve the issues associated with potential vapor space corrosion and liquid/air interface corrosion in the Type III highlevelwaste tanks is in place. The objective of the program is to develop understanding of vapor space (VSC) and liquid/air interface (LAIC) corrosion to ensure a defensible technical basis to provide accurate corrosion evaluations with regard to vapor space and liquid/air interface corrosion. The results of the FY05 experiments are presented here. The experiments are an extension of the previous research on the corrosion of tank steel exposed to simple solutions to corrosion of the steel when exposed to complex highlevelwaste simulants. The testing suggested that decanting and the consequent residual species on the tank wall is the predominant source of surface chemistry on the tank wall. The laboratory testing has shown that at the boundary conditions of the chemistry control program for solutions greater than 1M NaNO{sub 3}{sup -}. Minor and isolated pitting is possible within crevices in the vapor space of the tanks that contain stagnant dilute solution for an extended period of time, specifically when residues are left on the tank wall during decanting. Liquid/air interfacial corrosion is possible in dilute stagnant solutions, particularly with high concentrations of chloride. The experimental results indicate that Tank 50 would be most susceptible to the potential for liquid/air interfacial corrosion or vapor space corrosion, with Tank 49 and 41 following, since these tanks are nearest to the chemistry control boundary conditions. The testing continues to show that the combination of well-inhibited solutions and mill-scale sufficiently protect against pitting in the Type III tanks.

This is the ``seventh`` and final Quarterly Report under the scope of work for cooperative agreement between the Western Interstate Energy Board and the US Department of Energy. The report covers the period January--March 1993. The cooperative agreement was to expire in June 1992, but DOE granted an extension through March 24, 1993. Since this is the last Quarterly Report under the expired cooperative agreement, most tasks are noted as being completed. Two final items, however, will soon be sent to DOE -- final minutes from the March 9--11 High- Level Radioactive Waste Committee meeting, and the Year-End Technical Report. Some highlights from the quarter: The Committee decided on a preferred format for the revised Spent Fuel and High-Level Radioactive Waste Transportation Primer. The document would be 100- 200 pages, accompanied by a series of white papers on key transportation elements. A 25--30 page handbook for educating western state elected officials would also be prepared. On March 24, the Committee sent a letter to DOE commenting on the Near-Site Transportation Infrastructure report findings. The Committee is concerned that infrastructure limitations may limit the rail shipping option in many instances, even after upgrades have been implemented. The NSTI findings may also have significant relevance to the decision to develop multi-purpose canisters. On April 1, the Committee sent DOE the white paper, Transportation Implications of Various NWPA Program Options, which determined that DOE cannot develop a national transportation system by 1998 for shipments to an MRS or other federal storage facility.

A literature review was conducted to support the Task Technical and Quality Assurance Plan for Alternative Enhanced Chemical Cleaning (AECC) for sludge heel removal funded as part of the EM-21 Engineering and Technology program. The goal was to identify potential technologies or enhancements to the baseline oxalic acid cleaning process for chemically dissolving or mobilizing Savannah River Site (SRS) sludge heels. The issues with the potentially large volume of oxalate solids generated from the baseline process have driven an effort to find an improved or enhanced chemical cleaning technology for the tank heels. This literature review builds on a previous review conducted in 2003. A team was charged with evaluating the information in these reviews and developing recommendations of alternative technologies to pursue. The new information in this report supports the conclusion of the previous review that oxalic acid remains the chemical cleaning agent of choice for dissolving the metal oxides and hydroxides found in sludge heels in carbon steel tanks. The potential negative impact of large volumes of sodium oxalate on downstream processes indicates that the amount of oxalic acid used for chemical cleaning needs to be minimized as much as possible or the oxalic acid must be destroyed prior to pH adjustment in the receipt tank. The most straightforward way of minimizing the volume of oxalic acid needed for chemical cleaning is through more effective mechanical cleaning. Using a mineral acid to adjust the pH of the sludge prior to adding oxalic acid may also help to minimize the volume of oxalic acid used in chemical cleaning. If minimization of oxalic acid proves insufficient in reducing the volume of oxalate salts, several methods were found that could be used for oxalic acid destruction. For some waste tank heels, another acid or even caustic treatment (or pretreatment) might be more appropriate than the baseline oxalic acid cleaning process. Caustic treatment of high

This bibliography identifies documents that are pertinent to the design of waste packages for geologic disposal of nuclear waste. The bibliography is divided into fourteen subject categories so that anyone wishing to review the subject of leaching, for example, can turn to the leaching section and review the abstracts of reports which are concerned primarily with leaching. Abstracts are also cross referenced according to secondary subject matter so that one can get a complete list of abstracts for any of the fourteen subject categories. All documents which by their title alone appear to deal with the design of waste packages for the geologic disposal of spent fuel or high-levelwaste were obtained and reviewed. Only those documents which truly appear to be of interest to a waste package designer were abstracted. The documents not abstracted are listed in a separate section. There was no beginning date for consideration of a document for review. About 1100 documents were reviewed and about 450 documents were abstracted.

'U.C. Berkeley has made excellent progress in the last year in building and running experiments and performing analysis to study mixing processes that can affect the distribution of fuel and oxygen in the air space of DOE high-levelwaste tanks, and the potential to create flammable concentrations at isolated locations, achieving all of the milestones outlined in the proposal. The DOE support has allowed the acquisition of key experimental equipment, and has funded the full-time efforts of one doctoral student and one postdoctoral researcher working on the project. In addition, one masters student and one other doctoral student, funded by external sources, have also contributed to the research effort. Flammable gases can be generated in DOE high-levelwaste tanks, including radiolytic hydrogen, and during cesium precipitation from salt solutions, benzene. Under normal operating conditions the potential for deflagration or detonation from these gases is precluded by purging and ventilation systems, which remove the flammable gases and maintain a well-mixed condition in the tanks. Upon failure of the ventilation system, due to seismic or other events, however, it has proven more difficult to make strong arguments for well-mixed conditions, due to the potential for density-induced stratification which can potentially sequester fuel or oxidizer at concentrations significantly higher than average. This has complicated the task of defining the safety basis for tank operation. The author is currently developing numerical tools for modeling the transient evolution of fuel and oxygen concentrations in waste tanks following loss of ventilation. When used with reasonable grid resolutions, standard multi-dimensional fluid dynamics codes suffer from excessive numerical diffusion effects, which strongly over predict mixing and provide nonconservative estimates, particularly after stratification occurs. The National Institute of Standards and Technology (NIST) has developed

The U.S. Department of Energy's Office (DOE) of River Protection (ORP) has a continuing program for chemical optimization to better characterize corrosion behavior of High-LevelWaste (HLW). The DOE controls the chemistry in its HLW to minimize the propensity of localized corrosion, such as pitting, and stress corrosion cracking (SCC) in nitrate-containing solutions. By improving the control of localized corrosion and SCC, the ORP can increase the life of the Double-Shell Tank (DST) carbon steel structural components and reduce overall mission costs. The carbon steel tanks at the Hanford Site are critical to the mission of safely managing stored HLW until it can be treated for disposal. The DOE has historically used additions of sodium hydroxide to retard corrosion processes in HLW tanks. This also increases the amount of waste to be treated. The reactions with carbon dioxide from the air and solid chemical species in the tank continually deplete the hydroxide ion concentration, which then requires continued additions. The DOE can reduce overall costs for caustic addition and treatment of waste, and more effectively utilize waste storage capacity by minimizing these chemical additions. Hydroxide addition is a means to control localized and stress corrosion cracking in carbon steel by providing a passive environment. The exact mechanism that causes nitrate to drive the corrosion process is not yet clear. The SCC is less of a concern in the newer stress relieved double shell tanks due to reduced residual stress. The optimization of waste chemistry will further reduce the propensity for SCC. The corrosion testing performed to optimize waste chemistry included cyclic potentiodynamic volarization studies. slow strain rate tests. and stress intensity factor/crack growth rate determinations. Laboratory experimental evidence suggests that nitrite is a highly effective:inhibitor for pitting and SCC in alkaline nitrate environments. Revision of the corrosion control

High-level nuclear waste from the nuclear power industry and the nuclear weapons complex requires a geologic repository for disposal. The United States has identified a site at Yucca Mountain, Nevada for this purpose. Site selection and characterization have relied on input from earth scientists, including information about waste form behavior in a repository environment. The question of waste forms is central to the issue of reducing uncertainties in predicting the performance of a geologic repository over thousands to hundreds of thousands of years, the periods covered by standards that govern the release of radioactivity into the environment. The repository characterized by the U.S. Department of Energy at Yucca Mountain will be located in the unsaturated zone and will consequently provide an oxidizing environment for the waste emplaced there. The geochemical environment will have a large impact on the behavior of the waste, the majority of which will be spent nuclear fuel from nuclear power plants. Spent fuel is largely composed of UO2, in addition to a small fraction of fission products and actinides. Studies of natural analogues have shown that UO2 is not stable in an oxidizing environment in the presence of water and will alter to other phases. The ability of these alteration phases to contain radioactivity is still a subject of analysis. Other countries have addressed the issue of waste form stability by selecting a reducing environment in a location below the water table for their repositories. Such a choice increases the durability of spent fuel by orders of magnitude and thus reduces the uncertainties associated with predicting the performance of the repository. In the United States, the emphasis has turned away from the selection of appropriate natural barriers to transport of radioactivity and towards improvement of engineered barriers. This shift reflects a decision to abandon the natural barriers in favor of engineered ones, resulting in the

We have conducted a series of deliquescence, boiling point, chemical transformation, and evaporation experiments to determine the composition of waters likely to contact waste package surfaces over the thermal history of the repository as it heats up and cools back down to ambient conditions. In the above-boiling period, brines will be characterized by high nitrate to chloride ratios that are stable to higher temperatures than previously predicted. This is clearly shown for the NaCl-KNO{sub 3} salt system in the deliquescence and boiling point experiments in this report. Our results show that additional thermodynamic data are needed in nitrate systems to accurately predict brine stability and composition due to salt deliquescence in dust deposited on waste package surfaces. Current YMP models capture dry-out conditions but not composition for NaCl-KNO{sub 3} brines, and they fail to predict dry-out conditions for NaCl-KNO{sub 3}-NaNO{sub 3} brines. Boiling point and deliquescence experiments are needed in NaCl-KNO{sub 3}-NaNO{sub 3} and NaCl-KNO{sub 3}-NaNO{sub 3}-Ca(NO{sub 3}){sub 2} systems to directly determine dry-out conditions and composition, because these salt mixtures are also predicted to control brine composition in the above-boiling period. Corrosion experiments are needed in high temperature and high NO{sub 3}:Cl brines to determine if nitrate inhibits corrosion in these concentrated brines at temperatures above 160 C. Chemical transformations appear to be important for pure calcium- and magnesium-chloride brines at temperatures greater than 120 C. This stems from a lack of acid gas volatility in NaCl/KNO{sub 3} based brines and by slow CO{sub 2}(g) diffusion in alkaline brines. This suggests that YMP corrosion models based on bulk solution experiments over the appropriate composition, temperature, and relative humidity range can be used to predict corrosion in thin brine films formed by salt deliquescence. In contrast to the above-boiling period, the

The Nevada Test Site (NTS) region is located within the central section of a north-northeast-trending basaltic volcanic belt of late Cenozoic age, a part of the Quaternary volcanic province of the Great Basin. Future volcanism within the belt represents a potential hazard to storage of high-level radioactive waste within a buried repository located in the southwestern NTS. The hazards of future volcanism in the region are being characterized through a combination of volcanic hazards studies, probability determinations, and consequence analyses. Basaltic activity within the NTS regions is divided into two age groups consisting of relatively large-volume silicic cycle basalts (8 to 10 Myr) and rift basalts (< 8 to 0.3 Myr). This paper describes the processes of basaltic magmatism ranging from derivation of basalt melts at depth, through ascent through the upper mantle and crust, to surface eruption. Each stage in the evolution and dispersal of basaltic magma is described, and the disruption and potential dispersal of stored radioactive waste is evaluated. These data document areas of knowns and unknowns in the processes of basaltic volcanisms and provide background data necessary to assist calculations of radiation release levels due to disruption of a repository. 9 figures, 11 tables.

Radioactive highlevelwaste (HLW) sludge generated as a result of decades of production and manufacturing of plutonium, tritium and other nuclear materials is being removed from storage tanks and processed into a glass waste-form for permanent disposition at the Federal Repository. Characterization of this HLW sludge is a prerequisite for effective planning and execution of sludge disposition activities. The radioactivity of HLW makes sampling and analysis of the sludge very challenging, as well as making opportunities to perform characterization rare. In order to maximize the benefit obtained from sampling and analysis, a recommended list of physical property and chemical measurements has been developed. This list includes distribution of solids (insoluble and soluble) and water; densities of insoluble solids, interstitial solution, and slurry rheology (yield stress and consistency); mineral forms of solids; and primary elemental and radioactive constituents. Sampling requirements (number, type, volume, etc.), sample preparation techniques, and analytical methods are discussed in the context of pros and cons relative to end use of the data. Generation of useful sample identification codes and entry of results into a centralized database are also discussed.

Following completion of the preliminary risk assessment of the potential Yucca Mountain Site by Pacific Northwest Laboratory (PNL) in 1988, the Office of Civilian Radioactive Waste Management (OCRWM) of the US Department of Energy (DOE) requested the Performance Assessment Scientific Support (PASS) Program at PNL to develop an integrated system model and computer code that provides performance and risk assessment analysis capabilities for a potential high-level nuclear waste repository. The system model that has been developed addresses the cumulative radionuclide release criteria established by the US Environmental Protection Agency (EPA) and estimates population risks in terms of dose to humans. The system model embodied in the SUMO (System Unsaturated Model) code will also allow benchmarking of other models being developed for the Yucca Mountain Project. The system model has three natural divisions: (1) source term, (2) far-field transport, and (3) dose to humans. This document gives a detailed description of the mathematics of each of these three divisions. Each of the governing equations employed is based on modeling assumptions that are widely accepted within the scientific community.

Metallic engineered barriers must provide a period of absolute containment to high-level radioactive waste in geological repositories. Candidate materials include copper alloys, carbon steels, stainless steels, nickel alloys, and titanium alloys. The national programs of nuclear waste management have to identify and assess the anticipated degradation modes of the selected materials in the corresponding repository environment, which evolves in time. Commonly assessed degradation modes include general corrosion, localized corrosion, stress-corrosion cracking, hydrogen-assisted cracking, and microbiologically influenced corrosion. Laboratory testing and modeling in metallurgical and environmental conditions of similar and higher aggressiveness than those expected in service conditions are used to evaluate the corrosion resistance of the materials. This review focuses on the anticipated degradation modes of the selected or reference materials as corrosion-resistant barriers in nuclear repositories. These degradation modes depend not only on the selected alloy but also on the near-field environment. The evolution of the near-field environment varies for saturated and unsaturated repositories considering backfilled and unbackfilled conditions. In saturated repositories, localized corrosion and stress-corrosion cracking may occur in the initial aerobic stage, while general corrosion and hydrogen-assisted cracking are the main degradation modes in the anaerobic stage. Unsaturated repositories would provide an oxidizing environment during the entire repository lifetime. Microbiologically influenced corrosion may be avoided or minimized by selecting an appropriate backfill material. Radiation effects are negligible provided that a thick-walled container or an inner shielding container is used.

The generation of colloids in the interaction of high-level nuclear waste glasses with groundwater at 90{degrees}C has been investigated. The stability of the colloidal suspensions has been characterized with respect to salt concentration, pH time, particle size, and zeta potential. The compositions and the morphology of the colloids have also been determined with transmission electron microscopy (TEM). From ourtest results combined with earlier ones, we conclude that the waste glass may contribute to the colloid formation by increasing ion concentration in groundwater, which causes nucleation of colloids; by releasing radionuclides that adsorb onto existing groundwater colloids; and by spalling colloidal-size fragments from the surface layer of the reacted glass. The colloids are silicon-rich particles, such as smectites and uranium silicates. When the salt concentration in the solution is high the colloidal suspensions agglomerate. However, the agglomerated particles can be resuspended if the salt concentration is lowered by dilution with groundwater. The colloids agglomerate quickly after the leachate is cooled to room temperature. Most of the colloids settle out of the solution within a few days at ambient temperature. The isoelectric point is at a pH of approximately 1.0. Between pH 1 and 10.5, the colloids are negatively charged, which suggests that they will deposit readily on, positively charged surfaces. The average particle size islargest at the isoelectric point and is smallest around pH 6.

The generation of colloids in the interaction of high-level nuclear waste glasses with groundwater at 90[degrees]C has been investigated. The stability of the colloidal suspensions has been characterized with respect to salt concentration, pH time, particle size, and zeta potential. The compositions and the morphology of the colloids have also been determined with transmission electron microscopy (TEM). From ourtest results combined with earlier ones, we conclude that the waste glass may contribute to the colloid formation by increasing ion concentration in groundwater, which causes nucleation of colloids; by releasing radionuclides that adsorb onto existing groundwater colloids; and by spalling colloidal-size fragments from the surface layer of the reacted glass. The colloids are silicon-rich particles, such as smectites and uranium silicates. When the salt concentration in the solution is high the colloidal suspensions agglomerate. However, the agglomerated particles can be resuspended if the salt concentration is lowered by dilution with groundwater. The colloids agglomerate quickly after the leachate is cooled to room temperature. Most of the colloids settle out of the solution within a few days at ambient temperature. The isoelectric point is at a pH of approximately 1.0. Between pH 1 and 10.5, the colloids are negatively charged, which suggests that they will deposit readily on, positively charged surfaces. The average particle size islargest at the isoelectric point and is smallest around pH 6.

In the Defense Waste Processing Facility at the Savannah River Site, it is desired to eliminate a startup test that would involve adding small amounts of radioactive cesium-137 to simulated high-levelwaste. In order to eliminate this test, a reliable method for measuring non-radioactive cesium in the offgas system from the glass melter is required. From a pilot scale melter system, offgas particulate samples were taken on filter paper media and analyzed by Inductively Coupled Plasma-Mass Spectrometry (ICP-MS). The ICPMS method proved to be sufficiently sensitive to measure cesium quantities as low as 0.135 {mu}g, with the sensitivity being limited by the background cesium present in the filter paper. Typical particulate loadings ranged from <0.2 to >800 {mu}g of cesium. This sensitivity allowed determination of cesium decontamination factors for four of the five major components of the offgas system. The decontamination factors measured experimentally compared favorably with the process design basis values.

In the Defense Waste Processing Facility at the Savannah River Site, it is desired to eliminate a startup test that would involve adding small amounts of radioactive cesium-137 to simulated high-levelwaste. In order to eliminate this test, a reliable method for measuring non-radioactive cesium in the offgas system from the glass melter is required. From a pilot scale melter system, offgas particulate samples were taken on filter paper media and analyzed by Inductively Coupled Plasma-Mass Spectrometry (ICP-MS). The ICPMS method proved to be sufficiently sensitive to measure cesium quantities as low as 0.135 {mu}g, with the sensitivity being limited by the background cesium present in the filter paper. Typical particulate loadings ranged from <0.2 to >800 {mu}g of cesium. This sensitivity allowed determination of cesium decontamination factors for four of the five major components of the offgas system. The decontamination factors measured experimentally compared favorably with the process design basis values.

In 1985, the China National Nuclear Corporation (CNNC) worked out an R and D program called DG program for the deep geological disposal of high-level radioactive waste in China. The site selection process for China`s HLW repository has been carried out since then according to this program. Granite is considered as the candidate host rock for the repository. The general siting criteria are based on the principle that, under the effect of natural and human activities, the long term (100,000 years) safety of the repository can be reasonably obtained and the disposed radioactive waste can be avoided from entering the biosphere and harming human beings. During siting, two types of factors are considered: (1) social factors, including the nuclear industry distribution population, economic potential and environmental protection etc.; (2) natural factors, including geographic, meteorological and geological (crustal stability, host rocks, hydrogeology, engineering geology). The site selection process is divided into 4 stages: (1) nationwide screening, (2) regional screening; (3) district screening; and (4) site screening. During the first stage (1985--1986) the following were considered as potential regions: (1) southwest China, (2) Guangdong area, (3) Inner Mongolia, (4) east China and (5) northwest China. During the second stage (1986--1988), 21 districts were selected for further investigation. Since 1989 most efforts have been focused on the Beishan area, Gansu province, northwest China, which is considered as the most potential district for the repository.

This project plan has a two fold purpose. First, it provides a plan specific to the Hanford Tank Waste Remediation System (TWRS) Immobilized High-LevelWaste (EMW) Storage Subproject for the Washington State Department of Ecology (Ecology) that meets the requirements of Hanford Federal Facility Agreement and Consent Order (Tri-Party Agreement) milestone M-90-01 (Ecology et al. 1996) and is consistent with the project plan content guidelines found in Section 11.5 of the Tri-Party Agreement action plan. Second, it provides an upper tier document that can be used as the basis for future subproject line item construction management plans. The planning elements for the construction management plans are derived from applicable U.S. Department of Energy (DOE) planning guidance documents (DOE Orders 4700.1 (DOE 1992a) and 430.1 (DOE 1995)). The format and content of this project plan are designed to accommodate the plan`s dual purpose. A cross-check matrix is provided in Appendix A to explain where in the plan project planning elements required by Section 11.5 of the Tri-Party Agreement are addressed.

The Waste Vitrification Plant (WVP) converts Highly Active Liquor (HAL) from spent nuclear fuel reprocessing into a stable vitrified product. Recently WVP have been experiencing accumulation of solids in their primary off gas (POG) system leading to potential blockages. Chemical analysis of the blockage material via Laser Induced Breakdown Spectroscopy (LIBS) has shown it to exclusively consist of caesium, technetium and oxygen. The solids are understood to be caesium pertechnetate (CsTcO{sub 4}), resulting from the volatilisation of caesium and technetium from the highlevelwaste glass melt. Using rhenium as a chemical surrogate for technetium, a series of full scale experiments have been performed in order to understand the mechanism of rhenium volatilisation as caesium perrhenate (CsReO{sub 4}), and therefore technetium volatilisation as CsTcO{sub 4}. These experiments explored the factors governing volatilisation rates from the melt, potential methods of minimising the amount of volatilisation, and various strategies for mitigating the deleterious effects of the volatile material on the POG. This paper presents the results from those experiments, and discusses potential methods to minimise blockages that can be implemented on WVP, so that the frequency of the CsTcO{sub 4} blockages can be reduced or even eradicated altogether. (authors)

Large volumes of high-levelwaste (HLW) currently stored in tanks at DOE sites contain both sludges and supernatants. The sludges are composed of insoluble precipitates of actinides, radioactive fission products, and nonradioactive components. The supernatants are alkaline carbonate solutions, which can contain soluble actinides, fission products, metal ions, and high concentrations of major electrolytes including sodium hydroxide, nitrate, nitrite, phosphate, carbonate, aluminate, sulfate, and organic complexants. The organic complexants include several compounds that can form strong aqueous complexes with actinide species and fission products including ethylenediaminetetraacetic acid (EDTA), N-(2-hydroxyethyl)ethylenediaminetriacetic acid (HEDTA), nitrilotriacetic acid (NTA), iminodiacetic acid (IDA), citrate, glycolate, gluconate, and degradation products, formate and oxalate. The goal of this project is to determine the effects of hydrolysis, carbonate complexation, and metal ion displacement on trivalent and selected tetravalent actinide speciation in the presence of organic chelates present in tank waste and to use these data to develop accurate predictive thermodynamic models for use in chemical engineering applications at Hanford and other DOE sites.

The U.S. Department of Energy plans to vitrify approximately 60,000 metric tons of high-levelwaste (HLW) sludge from underground storage tanks at the Hanford Site in Southwest Washington State. To reduce the volume of HLW requiring treatment, a goal has been set to remove a significant quantity of the aluminum, which comprises nearly 70 percent of the sludge. Aluminum is found in the form of gibbsite and sodium aluminate, which can be easily dissolved by washing the waste stream with caustic, and boehmite, which comprises nearly half of the total aluminum, but is more resistant to caustic dissolution and requires higher treatment temperatures and hydroxide concentrations. Chromium, which makes up a much smaller amount ({approx}3%) of the sludge, must also be removed because there is a low tolerance for chromium in the HLW immobilization process. In this work, the coupled dissolution kinetics of aluminum and chromium species during caustic leaching of actual Hanford HLW samples is examined. The experimental results are used to develop a model that provides a basis for predicting dissolution dynamics from known process temperature and hydroxide concentration. (authors)

The ability to effectively mix, sample, certify, and deliver consistent batches of HighLevelWaste (HLW) feed from the Hanford Double Shell Tanks (DST) to the Waste Treatment and Immobilization Plant (WTP) presents a significant mission risk with potential to impact mission length and the quantity of HLW glass produced. DOE's River Protection Project (RPP) mission modeling and WTP facility modeling assume that individual 3785 cubic meter (l million gallon) HLW feed tanks are homogenously mixed, representatively sampled, and consistently delivered to the WTP. It has been demonstrated that homogenous mixing ofHLW sludge in Hanford DSTs is not likely achievable with the baseline design thereby causing representative sampling and consistent feed delivery to be more difficult. Inconsistent feed to the WTP could cause additional batch-to-batch operational adjustments that reduce operating efficiency and have the potential to increase the overall mission length. The Hanford mixing and sampling demonstration program will identify DST mixing performance capability, will evaluate representative sampling techniques, and will estimate feed batch consistency. An evaluation of demonstration program results will identify potential mission improvement considerations that will help ensure successful mission completion. This paper will discuss the history, progress, and future activities that will define and mitigate the mission risk.

Solubility testing with simulated HighLevelWaste tank heel solids has been conducted in order to evaluate two alternative chemical cleaning technologies for the dissolution of sludge residuals remaining in the tanks after the exhaustion of mechanical cleaning and sludge washing efforts. Tests were conducted with non-radioactive pure phase metal reagents, binary mixtures of reagents, and a Savannah River Site PUREX heel simulant to determine the effectiveness of an optimized, dilute oxalic/nitric acid cleaning reagent and pure, dilute nitric acid toward dissolving the bulk non-radioactive waste components. A focus of this testing was on minimization of oxalic acid additions during tank cleaning. For comparison purposes, separate samples were also contacted with pure, concentrated oxalic acid which is the current baseline chemical cleaning reagent. In a separate study, solubility tests were conducted with radioactive tank heel simulants using acidic and caustic permanganate-based methods focused on the “targeted” dissolution of actinide species known to be drivers for Savannah River Site tank closure Performance Assessments. Permanganate-based cleaning methods were evaluated prior to and after oxalic acid contact.

sludge solids. We recommend that these results be evaluated further to determine if these solutions contain sufficient neutron poisons. We observed low general corrosion rates in tests in which carbon steel coupons were contacted with solutions of oxalic acid, citric acid and mixtures of oxalic and citric acids. Wall thinning can be minimized by maintaining short contact times with these acid solutions. We recommend additional testing with oxalic and oxalic/citric acid mixtures to measure dissolution performance of sludges that have not been previously dried. This testing should include tests to clearly ascertain the effects of total acid strength and metal complexation on dissolution performance. Further work should also evaluate the downstream impacts of citric acid on the SRSHigh-LevelWaste System (e.g., radiochemical separations in the Salt Waste Processing Facility and addition of organic carbon in the Saltstone and Defense Waste Processing facilities).

The High-LevelWaste Storage Tank Farms/242-A Evaporator Standards/Requirements Identification Document (S/RID) is contained in multiple volumes. This document (Volume 6) outlines the standards and requirements for the sections on: Environmental Restoration and Waste Management, Research and Development and Experimental Activities, and Nuclear Safety.

Disposal of high-level radioactive waste, including spent nuclear fuel, in deep (3 to 5 km) boreholes is a potential option for safely isolating these wastes from the surface and near-surface environment. Existing drilling technology permits reliable and cost-effective construction of such deep boreholes. Conditions favorable for deep borehole disposal in crystalline basement rocks, including low permeability, high salinity, and geochemically reducing conditions, exist at depth in many locations, particularly in geologically stable continental regions. Isolation of waste depends, in part, on the effectiveness of borehole seals and potential alteration of permeability in the disturbed host rock surrounding the borehole. Coupled thermal-mechanical-hydrologic processes induced by heat from the radioactive waste may impact the disturbed zone near the borehole and borehole wall stability. Numerical simulations of the coupled thermal-mechanical response in the host rock surrounding the borehole were conducted with three software codes or combinations of software codes. Software codes used in the simulations were FEHM, JAS3D, Aria, and Adagio. Simulations were conducted for disposal of spent nuclear fuel assemblies and for the higher heat output of vitrified waste from the reprocessing of fuel. Simulations were also conducted for both isotropic and anisotropic ambient horizontal stress in the host rock. Physical, thermal, and mechanical properties representative of granite host rock at a depth of 4 km were used in the models. Simulation results indicate peak temperature increases at the borehole wall of about 30 C and 180 C for disposal of fuel assemblies and vitrified waste, respectively. Peak temperatures near the borehole occur within about 10 years and decline rapidly within a few hundred years and with distance. The host rock near the borehole is placed under additional compression. Peak mechanical stress is increased by about 15 MPa (above the assumed ambient

Disposal of high-level radioactive waste, including spent nuclear fuel, in deep (3 to 5 km) boreholes is a potential option for safely isolating these wastes from the surface and near-surface environment. Existing drilling technology permits reliable and cost-effective construction of such deep boreholes. Conditions favorable for deep borehole disposal in crystalline basement rocks, including low permeability, high salinity, and geochemically reducing conditions, exist at depth in many locations, particularly in geologically stable continental regions. Isolation of waste depends, in part, on the effectiveness of borehole seals and potential alteration of permeability in the disturbed host rock surrounding the borehole. Coupled thermal-mechanical-hydrologic processes induced by heat from the radioactive waste may impact the disturbed zone near the borehole and borehole wall stability. Numerical simulations of the coupled thermal-mechanical response in the host rock surrounding the borehole were conducted with three software codes or combinations of software codes. Software codes used in the simulations were FEHM, JAS3D, Aria, and Adagio. Simulations were conducted for disposal of spent nuclear fuel assemblies and for the higher heat output of vitrified waste from the reprocessing of fuel. Simulations were also conducted for both isotropic and anisotropic ambient horizontal stress in the host rock. Physical, thermal, and mechanical properties representative of granite host rock at a depth of 4 km were used in the models. Simulation results indicate peak temperature increases at the borehole wall of about 30 °C and 180 °C for disposal of fuel assemblies and vitrified waste, respectively. Peak temperatures near the borehole occur within about 10 years and decline rapidly within a few hundred years and with distance. The host rock near the borehole is placed under additional compression. Peak mechanical stress is increased by about 15 MPa (above the assumed ambient

A key aspect of high-level liquid-waste spray calcination is waste-feed atomization by using air atomizing nozzles. Atomization substantially increases the heat transfer area of the waste solution, which enhances rapid drying. Experience from the spray-calciner operations has demonstrated that nozzle flow conditions that produce 70-..mu.. median-volume-diameter or smaller spray droplets are required for small-scale spray calciners (drying capacity less than 80 L/h). For large-scale calciners (drying capacity greater than 300 L/h), nozzle flow conditions that produce 100-..mu.. median-volume-diameter or smaller spray droplets are required. Mass flow ratios of 0.2 to 0.4, depending on nozzle size, are required for proper operation of internal-mix atomizing nozzles. Both internal-mix and external-mix nozzles have been tested at PNL. Due to the lower airflow requirements and fewer large droplets produced, the internal-mix nozzle has been chosen for primary development in the spray calciner program at PNL. Several nozzle air-cap materials for internal-mix nozzles have been tested for wear resistance. Results show that nozzle air caps of stainless steel and Cer-vit (a machineable glass ceramic) are suceptible to rapid wear by abrasive slurries, whereas air caps of alumina and reaction-bonded silicon nitride show only slow wear. Longer-term testing is necessary to determine more accurately the actual frequency of nozzle replacement. Atomizing nozzle air caps of alumina are subject to fracture from thermal shock, whereas air caps of silicon nitride and Cer-vit are not. Fractured nozzles are held in place by the air-cap retaining ring and continue to atomize satisfactorily. Therefore, fractures caused by thermal shocking do not necessarily result in nozzle failure.

This document presents work conducted in support of technical needs expressed, in part, by the Engineering, Procurement, and Construction Contractor for the Salt Waste Processing Facility (SWPF). The Department of Energy (DOE) requested that Savannah River National Laboratory (SRNL) analyze and characterize supernate waste from seven selected HighLevelWaste (HLW) tanks to allow: classification of feed to be sent to the SWPF; verification that SWPF processes will be able to meet Saltstone Waste Acceptance Criteria (WAC); and updating of the Waste Characterization System (WCS) database. This document provides characterization data of samples obtained from Tanks 13H, 30H, 37H, 39H, 45F, 46F, and 49H and discusses results. Characterization of the waste tank samples involved several treatments and analysis at various stages of sample processing. These analytical stages included as-received liquid, post-dilution to 6.44 M sodium (target), post-acid digestion, post-filtration (at 3 filtration pore sizes), and after cesium removal using ammonium molybdophosphate (AMP). All tanks will require cesium removal as well as treatment with Monosodium Titanate (MST) for {sup 90}Sr (Strontium) decontamination. A small filtration effect for 90Sr was observed for six of the seven tank wastes. No filtration effects were observed for Pu (Plutonium), Np (Neptunium), U (Uranium), or Tc (Technetium); {sup 137}Cs (Cesium) concentration is ~E+09 pCi/mL for all the tank wastes. Tank 37H is significantly higher in {sup 90}Sr than the other six tanks. {sup 237}Np in the F-area tanks (45F and 46F) are at least 1 order of magnitude less than the H-Area tank wastes. The data indicate a constant ratio of {sup 99}Tc to Cs in the seven tank wastes. This indicates the Tc remains largely soluble in Savannah River Site (SRS) waste and partitions similarly with Cs. {sup 241}Am (Americium) concentration was low in the seven tank wastes. The majority of data were detection limit values, the largest being

This report provides a collection of annotated bibliographies for documents prepared under the Hanford High-LevelWaste Vitrification (Plant) Program. The bibliographies are for documents from Fiscal Year 1983 through Fiscal Year 1995, and include work conducted at or under the direction of the Pacific Northwest National Laboratory. The bibliographies included focus on the technology developed over the specified time period for vitrifying Hanford pretreated high-levelwaste. The following subject areas are included: General Documentation; Program Documentation; High-LevelWaste Characterization; Glass Formulation and Characterization; Feed Preparation; Radioactive Feed Preparation and Glass Properties Testing; Full-Scale Feed Preparation Testing; Equipment Materials Testing; Melter Performance Assessment and Evaluations; Liquid-Fed Ceramic Melter; Cold Crucible Melter; Stirred Melter; High-Temperature Melter; Melter Off-Gas Treatment; Vitrification Waste Treatment; Process, Product Control and Modeling; Analytical; and Canister Closure, Decontamination, and Handling

The legacy nuclear wastes stored in underground tanks at the US Department of Energy’s Hanford site is planned to be separated into high-levelwaste and low-activity waste fractions and vitrified separately. Formulating optimized glass compositions that maximize the waste loading in glass is critical for successful and economical treatment and immobilization of nuclear wastes. Glass property-composition models have been developed and applied to formulate glass compositions for various objectives for the past several decades. The property models with associated uncertainties and combined with composition and property constraints have been used to develop preliminary glass formulation algorithms designed for vitrification process control and waste form qualification at the planned waste vitrification plant. This paper provides an overview of current status of glass property-composition models, constraints applicable to Hanford waste vitrification, and glass formulation approaches that have been developed for vitrification of hazardous and highly radioactive wastes stored at the Hanford site.

The separation of trivalent actinides and rare earths from other fission products in high-level nuclear wastes by oxalate precipitation followed by ion exchange (OPIX) was experimentally investigated using synthetic wastes and a small-scale, continuous-flow oxalic acid precipitation and solid-liquid separation system. Trivalent actinide and rare earth oxalates are relatively insoluble in 0.5 to 1.0 M HNO/sub 3/ whereas other fission product oxalates are not. The continuous-flow system consisted of one or two stirred-tank reactors in series for crystal growth. Oxalic acid and waste solutions were mixed in the first tank, with the product solid-liquid slurry leaving the second tank. Solid-liquid separation was tested by filters and by a gravity settler. The experiments determined the fraction of rare earths precipitated and separated from synthetic waste streams as a function of number of reactors, system temperature, oxalic acid concentration, liquid residence time in the process, power input to the stirred-tank reactors, and method of solid-liquid separation. The crystalline precipitate was characterized with respect to form, size, and chemical composition. These experiments are only the first step in converting a proposed chemical flowsheet into a process flowsheet suitable for large-scale remote operations at high activity levels.

The papers in this volume cover the following subjects: waste isolation and the natural geohydrologic system; repository perturbations of the natural system; radionuclide migration through the natural system; and repository design technology. Individual papers are abstracted.

This report presents the results of a survey of the waste package strategies for seven western countries with active nuclear power programs that are pursuing disposal of spent nuclear fuel or high-levelwastes in deep geologic rock formations. Information, current as of January 1989, is given on the leading waste package concepts for Belgium, Canada, France, Federal Republic of Germany, Sweden, Switzerland, and the United Kingdom. All but two of the countries surveyed (France and the UK) have developed design concepts for their repositories, but none of the countries has developed its final waste repository or package concept. Waste package concepts are under study in all the countries surveyed, except the UK. Most of the countries have not yet developed a reference concept and are considering several concepts. Most of the information presented in this report is for the current reference or leading concepts. All canisters for the wastes are cylindrical, and are made of metal (stainless steel, mild steel, titanium, or copper). The canister concepts have relatively thin walls, except those for spent fuel in Sweden and Germany. Diagrams are presented for the reference or leading concepts for canisters for the countries surveyed. The expected lifetimes of the conceptual canisters in their respective disposal environment are typically 500 to 1,000 years, with Sweden's copper canister expected to last as long as one million years. Overpack containers that would contain the canisters are being considered in some of the countries. All of the countries surveyed, except one (Germany) are currently planning to utilize a buffer material (typically bentonite) surrounding the disposal package in the repository. Most of the countries surveyed plan to limit the maximum temperature in the buffer material to about 100{degree}C. 52 refs., 9 figs.

A source term model based on two-component leach flux concept is developed for a high-level radioactive waste repository. The long-lived radionuclides associated with high-levelwaste may give rise to the build-up of activity because of radioactive decay chains. The ingrowths of progeny are incorporated in the model using Bateman decay chain build-up equations. The model is applied to different radionuclides present in the high-level radioactive waste, which form a part of decay chains (4n to 4n + 3 series), and the activity of the parent and daughter radionuclides leaching out of the waste matrix is estimated. Two cases are considered: one when only parent is present initially in the waste and another where daughters are also initially present in the waste matrix. The incorporation of in situ production of daughter radionuclides in the source is important to carry out realistic estimates. It is shown that the inclusion of decay chain build-up is essential to avoid underestimation of the radiological impact assessment of the repository. The model can be a useful tool for evaluating the source term of the radionuclide transport models used for the radiological impact assessment of high-level radioactive waste repositories. PMID:27337157

This Environmental Assessment (EA) has been prepared by the Department of Energy (DOE) to assess the potential environmental impacts associated with the closure of 51 high-level radioactive waste tanks and tank farm ancillary equipment (including transfer lines, evaporators, filters, pumps, etc) at the Savannah River Site (SRS) located near Aiken, South Carolina. The waste tanks are located in the F- and H-Areas of SRS and vary in capacity from 2,839,059 liters (750,000 gallons) to 4,921,035 liters (1,300,000 gallons). These in-ground tanks are surrounded by soil to provide shielding. The F- and H-Area High-LevelWaste Tanks are operated under the authority of Industrial Wastewater Permits No.17,424-IW; No.14520, and No.14338 issued by the South Carolina Department of Health and Environmental Control (SCDHEC). In accordance with the Permit requirements, DOE has prepared a Closure Plan (DOE, 1996) and submitted it to SCDHEC for approval. The Closure Plan identifies all applicable or relevant and appropriate regulations, statutes, and DOE Orders for closing systems operated under the Industrial Wastewater Permits. When approved by SCDHEC, the Closure Plan will present the regulatory process for closing all of the F- and H-Area HighLevelWaste Tanks. The Closure Plan establishes performance objectives or criteria to be met prior to closing any tank, group of tanks, or ancillary tank farm equipment. The proposed action is to remove the residual wastes from the tanks and to fill the tanks with a material to prevent future collapse and bind up residual waste, to lower human health risks, and to increase safety in and around the tanks. If required, an engineered cap consisting of clay, backfill (soil), and vegetation as the final layer to prevent erosion would be applied over the tanks. The selection of tank system closure method will be evaluated against the following Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) criteria described in 40

Aluminum is a principal element in alkaline nuclear sludge waste stored in highlevelwaste (HLW) tanks at the Savannah River Site. The mass of sludge in a HLW tank can be reduced through the caustic leaching of aluminum, i.e. converting aluminum oxides (gibbsite) and oxide-hydroxides (boehmite) into soluble hydroxides through reaction with a hot caustic solution. The temperature limits outlined by the chemistry control program for HLW tanks to prevent caustic stress corrosion cracking (CSCC) in concentrated hydroxide solutions will potentially be exceeded during the sludge mass reduction (SMR) campaign. Corrosion testing was performed to determine the potential for CSCC under expected conditions. The experimental test program, developed based upon previous test results and expected conditions during the current SMR campaign, consisted of electrochemical and mechanical testing to determine the susceptibility of ASTM A516 carbon steel to CSCC in the relevant environment. Anodic polarization test results indicated that anodic inhibition at the temperatures and concentrations of interest for SMR is not a viable, consistent technical basis for preventing CSCC. However, the mechanical testing concluded that CSCC will not occur under conditions expected during SMR for a minimum of 35 days. In addition, the stress relief for the Type III/IIIA tanks adds a level of conservatism to the estimates. The envelope for corrosion control is recommended during the SMR campaign is shown in Table 1. The underlying assumption is that solution time-in-tank is limited to the SMR campaign. The envelope recommends nitrate/aluminate intervals for discrete intervals of hydroxide concentrations, although it is recognized that a continuous interval may be developed. The limits also sets temperature limits.

Three copper-based alloys, CDA 102 (oxygen-free, high-purity copper), CDA 613 (aluminum bronze), and CDA 715 (Cu-30Ni), are candidates for the fabrication of high-level radioactive-waste disposal containers. Waste will include spent fuel assemblies from reactors as well as borosilicate glass, and will be sent to the prospective repository site at Yucca Mountain in Nye County, Nevada. The decay of radionuclides will result in the generation of substantial heat and in fluxes of gamma radiation outside the containers. In this environment, container materials might degrade by atmospheric oxidation, general aqueous phase corrosion, localized corrosion (LC), and stress corrosion cracking (SCC). This volume is a critical survey of available data on pitting and crevice corrosion of the copper-based candidates. Pitting and crevice corrosion are two of the most common forms of LC of these materials. Data on the SCC of these alloys is surveyed in Volume 4. Pitting usually occurs in water that contains low concentrations of bicarbonate and chloride anions, such as water from Well J-13 at the Nevada Test Site. Consequently, this mode of degradation might occur in the repository environment. Though few quantitative data on LC were found, a tentative ranking based on pitting corrosion, local dealloying, crevice corrosion, and biofouling is presented. CDA 102 performs well in the categories of pitting corrosion, local dealloying, and biofouling, but susceptibility to crevice corrosion diminishes its attractiveness as a candidate. The cupronickel alloy, CDA 715, probably has the best overall resistance to such localized forms of attack. 123 refs., 11 figs., 3 tabs.

A numerical simulation framework is presented in this paper for estimating evolution of pH and release of major species from grout within high-levelwaste tanks after closure. This model was developed as part of the Cementitious Barriers Partnership. The reactive transport model consists of two parts - (1) transport of species, and (2) chemical reactions. The closure grout can be assumed to have varying extents of cracking and composition for performance assessment purposes. The partially or completely degraded grouted tank is idealized as a dual regime system comprising of a mobile region having solid materials with cracks and macro-pores, and an immobile/stagnant region having solid matrix with micropores. The transport profiles of the species are calculated by incorporating advection of species through the mobile region, diffusion of species through the immobile/stagnant region, and exchange of species between the mobile and immobile regions. A geochemical speciation code in conjunction with the pH dependent test data for a grout material is used to obtain a mineral set that best describes the trends in the test data of the major species. The dual regime reactive transport model predictions are compared with the release data from an up-flow column percolation test. The coupled model is then used to assess effects of crack state of the structure, rate and composition of the infiltrating water on the pH evolution at the grout-waste interface. The coupled reactive transport model developed in this work can be used as part of the performance assessment process for evaluating potential risks from leaching of a cracked tank containing elements of human health and environmental concern. (authors)

Gibbsite [Al(OH)(3)] and boehmite (AlOOH) have long been assumed to be the most prevalent aluminum-bearing minerals in Hanford high-level nuclear waste sludge. The present study shows that dawsonite [NaAl(OH)(2)CO(3)] is also a common aluminum-bearing phase in tanks containing high total inorganic carbon (TIC) concentrations and (relatively) low dissolved free hydroxide concentrations. Tank samples were probed for dawsonite by X-ray Diffraction (XRD), Scanning Electron Microscopy with Energy Dispersive Spectrometry (SEM-EDS) and Polarized Light Optical Microscopy. Dawsonite was conclusively identified in four of six tanks studied. In a fifth tank (AN-102), the dawsonite identification was less conclusive because it was only observed as a Na-Al bearing phase with SEM-EDS. Four of the five tank samples with dawsonite also had solid phase Na(2)CO(3) · H(2)O. The one tank without observable dawsonite (Tank C-103) had the lowest TIC content of any of the six tanks. The amount of TIC in Tank C-103 was insufficient to convert most of the aluminum to dawsonite (Al:TIC mol ratio of 20:1). The rest of the tank samples had much lower Al:TIC ratios (between 2:1 and 0.5:1) than Tank C-103. One tank (AZ-102) initially had dawsonite, but dawsonite was not observed in samples taken 15 months after NaOH was added to the tank surface. When NaOH was added to a laboratory sample of waste from Tank AZ-102, the ratio of aluminum to TIC in solution was consistent with the dissolution of dawsonite. The presence of dawsonite in these tanks is of significance because of the large amount of OH(-) consumed by dawsonite dissolution, an effect confirmed with AZ-102 samples. PMID:22326826

The present investigation aims to study the effect of various ions present in nuclear high-levelwaste (HLW) (acidic) medium on the corrosion resistance of Alloy 625, with solution-annealed and sensitized microstructure. The heat-affected zones are prone to sensitization during welding of components and subsequent exposure to acidic waste during service could result in intergranular corrosion in these regions and hence it was attempted to study the corrosion behavior of the alloy under sensitized conditions. Double loop electrochemical potentiokinetic reactivation test was carried out to obtain the extent of chromium depletion. Potentiodynamic anodic polarization and electrochemical noise investigations were carried out on Alloy 625 in 3 M nitric acid and simulated nuclear HLW medium (prepared in 3 M nitric acid) at 298 K and 323 K. The study showed that the alloy possess good corrosion resistance in 3 M nitric acid and simulated HLW medium. However, a marginal decrease in the corrosion resistance occurred in simulated HLW when compared to the plain acid, as observed from an increase in passivation current density, decrease in transpassive potentials, and decrease in electrochemical noise resistance. Increase in temperature of the medium and change in microstructure from solution-annealed to sensitized state further decreased the corrosion resistance of Alloy 625. Electrochemical noise time records obtained at open circuit conditions showed a stable passive film for 22 h of immersion of the alloy in 3 M nitric acid and simulated HLW. However, the amplitude of current fluctuations was higher for the sensitized microstructure when compared to the solution-annealed microstructure.

Controls on the solution chemistry (minimum nitrite and hydroxide concentrations) are in place to prevent the initiation and propagation of pitting and stress corrosion cracking in highlevelwaste (HLW) tanks. These controls are based upon a series of experiments performed on carbon steel coupons in simulated waste solutions. An experimental program was undertaken to investigate reducing the minimum molar nitrite concentration required to confidently inhibit pitting. A statistical basis to quantify the probability of pitting for the tank wall, when exposed to various dilute solutions, is being developed. Electrochemical and coupon testing are being performed within the framework of the statistical test matrix to determine the minimum necessary inhibitor concentrations and develop a quantitative model to predict pitting propensity. A subset of the original statistical test matrix was used to develop an applied understanding of the corrosion response of the carbon steel in the various environments. The interim results suggest that there exists some critical nitrite concentration that sufficiently inhibits against localized corrosion mechanisms due to nitrates/chlorides/sulfates, beyond which further nitrite additions are unnecessary. The combination of visual observation and the cyclic potentiodynamic polarization scans indicate the potential for significant inhibitor reductions without consequence specifically at nitrate concentrations near 1 M. The complete data sets will be used to determine the statistical basis to confidently inhibit against pitting using nitrite inhibition with the current pH controls. Once complete, a revised chemistry control program will be devised based upon the probability of pitting specifically for dilute solutions which will allow for tank specific chemistry control implementation.

The purpose of this report is to describe the expected environments associated with high-levelwaste (HLW) and spent fuel (SF) repositories in salt formations. These environments include the thermal, fluid, pressure, brine chemistry, and radiation fields predicted for the repository conceptual designs. In this study, it is assumed that the repository will be a room and pillar mine in a rock-salt formation, with the disposal horizon located approx. 2000 ft (610 m) below the surface of the earth. Canistered waste packages containing HLW in a solid matrix or SF elements are emplaced in vertical holes in the floor of the rooms. The emplacement holes are backfilled with crushed salt or other material and sealed at some later time. Sensitivity studies are presented to show the effect of changing the areal heat load, the canister heat load, the barrier material and thickness, ventilation of the storage room, and adding a second row to the emplacement configuration. The calculated thermal environment is used as input for brine migration calculations. The vapor and gas pressure will gradually attain the lithostatic pressure in a sealed repository. In the unlikely event that an emplacement hole will become sealed in relatively early years, the vapor space pressure was calculated for three scenarios (i.e., no hole closure - no backfill, no hole closure - backfill, and hole closure - no backfill). It was assumed that the gas in the system consisted of air and water vapor in equilibrium with brine. A computer code (REPRESS) was developed assuming that these changes occur slowly (equilibrium conditions). The brine chemical environment is outlined in terms of brine chemistry, corrosion, and compositions. The nuclear radiation environment emphasized in this report is the stored energy that can be released as a result of radiation damage or crystal dislocations within crystal lattices.

Ion exchange is recognized as an integral, if underrepresented, mechanism influencing glass corrosion. However, due to the formation of various alteration layers in the presence of water, it is difficult to conclusively deconvolute the mechanisms of ion exchange from other processes occurring simultaneously during corrosion. In this work, an operationally inert non-aqueous solution was used as an alkali source material to isolate ion exchange and study the solid-state diffusion of lithium. Specifically, the experiments involved contacting glass coupons relevant to the immobilization of high-level nuclear waste, SON68 and CJ-6, which contained Li in natural isotope abundance, with a non-aqueous solution of 6LiCl dissolved in dimethyl sulfoxide at 90 °C for various time periods. The depth profiles of major elements in the glass coupons were measured using time-of-flight secondary ion mass spectrometry (ToF-SIMS). Lithium interdiffusion coefficients, DLi, were then calculated based on the measured depth profiles. The results indicate that the penetration of 6Li is rapid in both glasses with the simplified CJ-6 glass (D6Li ≈ 4.0-8.0 × 10-21 m2/s) exhibiting faster exchange than the more complex SON68 glass (DLi ≈ 2.0-4.0 × 10-21 m2/s). Additionally, sodium ions present in the glass were observed to participate in ion exchange reactions; however, different diffusion coefficients were necessary to fit the diffusion profiles of the two alkali ions. Implications of the diffusion coefficients obtained in the absence of alteration layers to the long-term performance of nuclear waste glasses in a geological repository system are also discussed.

processing strategy for the Hanford Tank Waste Treatment and Immobilization Plant (WTP). The basis of this alternative approach is an empirical model predicting the crystal accumulation in the WTP glass discharge riser and melter bottom as a function of glass composition, time, and temperature. When coupled with an associated operating limit (e.g., the maximum tolerable thickness of an accumulated layer of crystals), this model could then be integrated into the process control algorithms to formulate crystal tolerant highlevelwaste (HLW) glasses targeting higher waste loadings while still meeting process related limits and melter lifetime expectancies. This report provides a review of the scaled melter testing that was completed in support of the Defense Waste Processing Facility (DWPF) melter. Testing with scaled melters provided the data to define the DWPF operating limits to avoid bulk (volume) crystallization in the un-agitated DWPF melter and provided the data to distinguish between spinels generated by K-3 refractory corrosion versus spinels that precipitated from the HLW glass melt pool. This report includes a review of the crystallization observed with the scaled melters and the full scale DWPF melters (DWPF Melter 1 and DWPF Melter 2). Examples of actual DWPF melter attainment with Melter 2 are given. The intent is to provide an overview of lessons learned, including some example data, that can be used to advance the development and implementation of an empirical model and operating limit for crystal accumulation for WTP. Operation of the first and second (current) DWPF melters has demonstrated that the strategy of using a liquidus temperature predictive model combined with a 100 °C offset from the normal melter operating temperature of 1150 °C (i.e., the predicted liquidus temperature (TL) of the glass must be 1050 °C or less) has been successful in preventing any detrimental accumulation of spinel in the DWPF melt pool, and spinel has not been

Investigate particles produced by laser irradiation and their analysis by Laser Ablation Inductively Coupled Plasma Mass Spectroscopy (LA/ICP-MS), with a view towards optimizing particle production for analysis of highlevelwaste materials and waste glass. LA/ICP-MS has considerable potential to increase the safety and speed of analysis required for the remediation of highlevelwastes from cold war plutonium production operations. In some sample types, notably the sodium nitrate-based wastes at Hanford and elsewhere, chemical analysis using typical laser conditions depends strongly on the details of sample history composition in a complex fashion, rendering the results of analysis uncertain. Conversely, waste glass materials appear to be better behaved and require different strategies to optimize analysis.

The twenty papers in this volume are divided into three parts: site exploration and characterization; repository development and design; and waste package development and design. These papers represent the status of technology that existed in 1981 and 1982. Individual papers were processed for inclusion in the Energy Data Base.

Solubility of noble metal elements (NME) in the melted borosilicate glass is much smaller than its normal concentration of the highlevel liquid waste. Thus most of NME show small particles in the melted glass and tend to sediment in the bottom region of the vitrification melter due to their higher density than that of glass. Experiments of the sedimentation of NME particles in the melted glass were carried out under static condition. Three conditions of initial NME concentration (1.1, 3.0, 6.1 wt % with an equivalent for each oxide) in the simulated glass were set and held at 1100 C. degrees up to 2880 hours. The specimen with 1.1 wt % initial NME concentration indicated zone settling, and the settling rate of the interface is constant: 2.4 mm/h. This sedimentation behavior is the type of rapid settling. Following the rapid settling, the settling rate goes gradually slower; this is the type of compressive settling. The specimens with 3.0 wt % and 6.1 wt % initial NME concentration showed compression settling from the beginning. From the settling curve of the interface, the maximum concentration of NME in sediment was estimated to be around 23- 26 wt %. Growth of NME particles was observed by holding at 1100 C. degrees for up to 2880 hours. The viscosity becomes higher as NME concentration increases and the dependence on shear rate becomes simultaneously stronger. The effect of the particle growth to viscosity appears to be not significant.

The dissolution of quartz particles and the growth and dissolution of crystalline phases during the conversion of batch to glass potentially affects both the glass melting process and product quality. Crystals of spinel exiting the cold cap to molten glass below can be troublesome during the vitrification of iron-containing high-levelwastes. To estimate the distribution of quartz and spinel fractions within the cold cap, we used kinetic models that relate fractions of these phases to temperature and heating rate. Fitting the model equations to data showed that the heating rate, apart from affecting quartz and spinel behavior directly, also affects them indirectly via concurrent processes, such as the formation and motion of bubbles. Because of these indirect effects, it was necessary to allow one kinetic parameter (the pre-exponential factor) to vary with the heating rate. The resulting kinetic equations are sufficiently simple for the detailed modeling of batch-to-glass conversion as it occurs in glass melters. The estimated fractions and sizes of quartz and spinel particles as they leave the cold cap, determined in this study, will provide the source terms needed for modeling the behavior of these solid particles within the flow of molten glass in the melter.

The aim of this study is to investigate the high-levelwaste (HLW) transmutation potential of fusion-driven transmuter (FDT) based on catalyzed D-D fusion plasma for various fuel fractions. The Minor actinide (MA) (237Np, 241Am, 243Am and 244Cm) and long-lived fission product (LLFP) (99Tc, 129I and 135Cs) nuclides discharged from high burn-up pressured water reactor-mixed oxide spent fuel are considered as the HLW. The volume fractions of the MA and LLFP are raised from 10 to 20% stepped by 2% and 10 to 80% stepped by 5%, respectively. The transmutation analyses have been performed for an operation period (OP) of up to 6 years by 75% plant factor ( η) under a first-wall neutron load ( P) of 5 MW/m2 by using two different computer codes, the XSDRNPM/SCALE4.4a neutron transport code and the MCNP4B Monte Carlo code. The numerical results bring out that the considered FDT has a high neutronic performance for an effective and rapid transmutation of MA and LLFP as well as the energy generation along the OP.

The primary goal of the US Subseabed Disposal Program (SDP) is to assess the technical and environmental feasibility of disposing of high-level nuclear waste in deep-sea sediments. The subseabed biology program is charged with assessing possible ecosystem effects of radionuclides as well as possible health effects to man from radionuclides which may be released in the deep sea and transported to the ocean surface. Current biological investigations are attempting to determine benthic community structure; benthic community metabolism; the biology of deep-sea mobile scavengers; the faunal composition of midwater nekton; rates of microbial processes; and the radiation sensitivity of deep-sea organisms. Existing models of the dispersal of radionuclides in the deep sea have not considered many of the possible biological mechanisms which may influence the movement of radionuclides. Therefore, a multi-compartment foodweb model is being developed which considers both biological and physical influences on radionuclide transport. This model will allow parametric studies to be made of the impact on the ocean environment and on man of potential releases of radionuclides.

Several countries favor the encapsulation of high-level radioactive waste (HLRW) in iron or steel canisters surrounded by highly compacted bentonite. In the present study the corrosion of iron in contact with different bentonites was investigated. The corrosion product was a 1:1 Fe layer silicate already described in literature (sometimes referred to as berthierine). Seven exposition test series (60 °C, 5 months) showed slightly less corrosion for the Na-bentonites compared to the Ca-bentonites. Two independent exposition tests with iron pellets and 38 different bentonites clearly proved the role of the layer charge density of the swelling clay minerals (smectites). Bentonites with high charged smectites are less corrosive than bentonites dominated by low charged ones. The type of counterion is additionally important because it determines the density of the gel and hence the solid/liquid ratio at the contact to the canister. The present study proves that the integrity of the multibarrier-system is seriously affected by the choice of the bentonite buffer encasing the metal canisters in most of the concepts. In some tests the formation of a patina was observed consisting of Fe-silicate. Up to now it is not clear why and how the patina formed. It, however, may be relevant as a corrosion inhibitor. PMID:25536393

This report is a preliminary attempt to plan a comprehensive program of geochemistry research aimed at resolving problems connected with the underground storage of high-level nuclear waste. The problems and research needs were identified in a companion report to this one. The research needs were taken as a point of departure and developed into a series of proposed projects with estimated manpowers and durations. The scope of the proposed research is based on consideration of an underground repository as a multiple barrier system. However, the program logic and organization reflect conventional strategies for resolving technological problems. The projects were scheduled and the duration of the program, critical path projects and distribution of manpower determined for both full and minimal programs. The proposed research was then compared with ongoing research within DOE, NRC and elsewhere to identify omissions in current research. Various options were considered for altering the scope of the program, and hence its cost and effectiveness. Finally, recommendations were made for dealing with omissions and uncertainties arising from program implementation. 11 references, 6 figures, 4 tables.

The West Valley Demonstration Project (WVDP) vitrification melter was shut down in September 2002 after being used to vitrify HighLevelWaste (HLW) and process system residuals for six years. Processing of the HLW occurred from June 1996 through November 2001, followed by a program to flush the remaining HLW through to the melter. Glass removal and shutdown followed. The facility and process equipment is currently in a standby mode awaiting deactivation. During HLW processing operations, nearly 24 million curies of radioactive material were vitrified into 275 canisters of HLW glass. At least 99.7% of the curies in the HLW tanks at the WVDP were vitrified using the melter. Each canister of HLW holds approximately 2000 kilograms of glass with an average contact dose rate of over 2600 rem per hour. After vitrification processing ended, two more cans were filled using the Evacuated Canister Process to empty the melter at shutdown. This history briefly summarizes the initial stages of process development and earlier WVDP experience in the design and operation of the vitrification systems, followed by a more detailed discussion of equipment availability and failure rates during six years of operation. Lessons learned operating a system that continued to function beyond design expectations also are highlighted.

The efficient storage and processing of high-level nuclear waste could be improved by a better understanding of the behaviour of the particle-laden fluid flows involved. This work reports a mathematical modeling study of impinging single and two-phase turbulent jets that is of relevance to the flows used industrially to prevent the settling of solid particles in storage tanks, and to re-suspend particles that form a bed. A computational fluid dynamic model, that embodies a Lagrangian particle tracking technique, is applied to the prediction of these flows. Predictions in the free flow and wall regions, and along the stagnation line, of the single phase flow are in reasonable accord with data, although the addition of particles results in less satisfactory agreement between predictions and measurements. The influence of particles is, however, reproduced qualitatively by the mathematical model, with quantitative differences attributable to a lack of particle drag in the simulations. Uncertainties in experimental parameters may be responsible for some of the differences between predictions and data, and examination of the data used casts doubts on its reliability. Further work is required in terms of the use of more advanced turbulence modeling techniques, and the provision of detailed and reliable data sets. (authors)

A key component of the Engineered Barrier System (EBS) being designed for containment of spent-fuel and high-levelwaste at the proposed geological repository at Yucca Mountain, Nevada is a two-layer canister. In this particular design, the inner barrier is made of a corrosion resistant material (CRM) such as Alloy 625 or C-22, while the outer barrier is made of a corrosion-allowance material (CAM) such as carbon steel or Monel400. Initially, the containers will be hot and dry due to the heat generated by radioactive decay. However, the temperature will eventually drop to levels where both humid air and aqueous phase corrosion will be possible. As the outer barrier is penetrated, uniform corrosion of the CRM will be possible in exfoliated areas. The possibility for crevice formation between the CAM and CRM will also exist. In the case of either Alloy 625 or C-22, a crevice will have to form before significant penetration of the CRM can occur. Crevice corrosion of the CRMs has been well documented. Lillard and Scully have induced crevice corrosion in Alloy 625 during exposure to artificial sea water. Jones and Wilde have prepared simulated crevice solutions of FeCl{sub 2}, NiCl{sub 2} and CrCl{sub 3}, and measured substantial pH suppression. Asphahani measured the dissolution rates of Alloys 625 and C-22 in such artificial crevice solutions at various temperatures. Others have observed no significant localized attack in less severe environments.

Under the sponsorship of the US Nuclear Regulatory Commission (NRC), Sandia National Laboratories (SNL) is developing a performance assessment methodology for the analysis of long-term disposal and isolation of high-level nuclear wastes (HLW) in alternative geologic media. As part of this exercise, SNL created a conceptualization of ground-water flow and radionuclide transport in the far field of a hypothetical HLW repository site located in unsaturated, fractured tuff formations. This study provides a foundation for the development of conceptual mathematical, and numerical models to be used in this performance assessment methodology. This conceptualization is site specific in terms of geometry, the regional ground-water flow system, stratigraphy, and structure in that these are based on information from Yucca Mountain located on the Nevada Test Site. However, in terms of processes in unsaturated, fractured, porous media, the model is generic. This report also provides a review and evaluation of previously proposed conceptual models of unsaturated and saturated flow and solute transport. This report provides a qualitative description of a hypothetical HLW repository site in fractured tuff. However, evaluation of the current knowledge of flow and transport at Yucca Mountain does not yield a single conceptual model. Instead, multiple conceptual models are possible given the existing information.

Strontium-90 is considered the most important radioactive isotope in the environment and one of the most frequently occurring radionuclides in groundwaters at nuclear facilities. The uranyl peroxide studtite (UO2O2 . 4H2O) has been observed to be formed in spent nuclear fuel leaching experiments and seems to have a relatively high sorption capacity for some radionuclides. In this work, the sorption of strontium onto studtite is studied as a function of time, strontium concentration in solution and pH. The main results obtained are (a) sorption is relatively fast although slower than for cesium; (b) strontium seems to be sorbed via a monolayer coverage of the studtite surface, (c) sorption has a strong dependence on ionic strength, is negligible at acidic pH, and increases at neutral to alkaline pH (almost 100% of the strontium in solution is sorbed above pH 10). These results point to uranium secondary solid phase formation on the spent nuclear fuel as an important mechanism for strontium retention in a high-level nuclear waste repository (HLNW). PMID:20638966

A key component of the Engineered Barrier System (EBS) being designed for containment of spent-fuel and high-levelwaste at the proposed geological repository at Yucca Mountain, Nevada is a two-layer canister. In this particular design, the inner barrier is made of a corrosion resistant material (CRM) such as Alloy 825, 625 or C-22, while the outer barrier is made of a corrosion-allowance material (CAM) such as A5 16 or Monel 400. At the present time, Alloy C-22 and A516 are favored. This publication addresses the development of models to account for corrosion of Alloy C-22 surfaces exposed directly to the Near Field Environment (NFE), as well as to the exacerbated conditions in the CAM-CRM crevice. [5]. Haynes International has published corrosion rates of Alloys 625 and C-22 in artificial crevice solutions (5-10 wt. % FeCl,) at various temperatures (25, 50 and 75 C) [6,7]. In this case, the observed rates for Alloy C-22 appear to be due to passive dissolution. It is believed that Alloy C-22 must be at an electrochemical potential above the repassivation potential to initiate localized corrosion.

A key component of the Engineered Barrier System (EBS) being designed for containment of spent-fuel and high-levelwaste at the proposed geological repository at Yucca Mountain, Nevada is a two-layer canister. In this particular design,the inner barrier is made of a corrosion resistant material (CRM) such as Alloy 625 and C-22, while the outer barrier is made of a corrosion allowance material (CAM) such as carbon steel or Monel 400. Initially, the containers will be hot and dry due to the heat generated by radioactive decay. However, the temperature will eventually drop to levels where both humid air and aqueous phase corrosion will be possible. As the outer barrier is penetrated, uniform corrosion of the CRM will be possible of exfoliated areas. The possibility of crevice formation between the CAM and CRM will also exist. In the case of either Alloy 625 or C-22, a crevice will have to form before significant penetration of the CRM can occur. Crevice corrosion of the CRMs has been well documented.

Transferring high-levelwaste (HLW) between storage tanks or to treatment facilities is a common practice performed at the Department of Energy (DoE) sites. Changes in the chemical and/or physical properties of the HLW slurry during the transfer process may lead to the formation of blockages inside the pipelines resulting in schedule delays and increased costs. To improve DoE's capabilities in the event of a pipeline plugging incident, FIU has continued to develop two novel unplugging technologies: an asynchronous pulsing system and a peristaltic crawler. The asynchronous pulsing system uses a hydraulic pulse generator to create pressure disturbances at two opposite inlet locations of the pipeline to dislodge blockages by attacking the plug from both sides remotely. The peristaltic crawler is a pneumatic/hydraulic operated crawler that propels itself by a sequence of pressurization/depressurization of cavities (inner tubes). The crawler includes a frontal attachment that has a hydraulically powered unplugging tool. In this paper, details of the asynchronous pulsing system's ability to unplug a pipeline on a small-scale test-bed and results from the experimental testing of the second generation peristaltic crawler are provided. The paper concludes with future improvements for the third generation crawler and a recommended path forward for the asynchronous pulsing testing. (authors)

The metering pump system that delivers high-level liquid wastes (HLLW) slurry to a melter is an integral subsystem of the vitrification process. The process of selecting a pump for this application began with a technical review of pumps typically used for slurry applications. The design and operating characteristics of numerous pumps were evaluated against established criteria. Two pumps, an air-displacement slurry (ADS) pump and an air-lift pump, were selected for further development. In the development activity, from FY 1983 to FY 1985, the two pumps were subjected to long-term tests using simulated melter feed slurries to evaluate the pumps' performances. Throughout this period, the designs of both pumps were modified to better adapt them for this application. Final reference designs were developed for both the air-displacement slurry pump and the air-lift pump. Successful operation of the final reference designs has demonstrated the feasibility of both pumps. A fully remote design of the ADS pump has been developed and is currently undergoing testing at the West Valley Demonstration Project. Five designs of the ADS pump were tested and evaluated. The initial four designs proved the operating concept of the ADS pump. Weaknesses in the ADS pump system were identified and eliminated in later designs. A full-scale air-lift pump was designed and tested as a final demonstration of the air-lift pump's capabilities.

An international consensus has emerged that deep geological disposal on land is one of the most appropriate means for highlevel radioactive wastes (HLW). The fluid transport is slow and radioactive elements are dangerous, so it's impossible to experiment over thousands of years. Instead, numerical model in such natural barrier as fractured underground needs to be considered. Field observations reveal that the equation with fractional derivative is more appropriate for describing physical phenomena than the equation which is based on the Fick's law. Thus, non-Fickian diffusion into inhomogeneous underground appears to be important in the assessment of HLW disposal. A solute transport equation with fractional derivative has been suggested and discussed in literature. However, no attempts were made to apply this equation for modeling of HLW disposal with account for the radioactive decay. In this study, we suggest the use of a novel fractional advection-diffusion equation which accounts for the effect of radioactive disintegration and for interactions between major, macro pores and fractal micro pores. This model is fundamentally different from previous proposed model of HLW, particularly in utilizing fractional derivative. Breakthrough curves numerically obtained by the present model are presented for a variety of rock types with respect to some important nuclides. Results of the calculation showed that for longer distance our model tends to be more conservative than the conventional Fickian model, therefore our model can be said to be safer.

An international consensus has emerged that deep geological disposal on land is one of the most appropriate means for highlevel radioactive wastes (HLW). The fluid transport is slow and radioactive elements are dangerous, so it's impossible to experiment over thousands of years. Instead, numerical model in such natural barrier as fractured underground needs to be considered. Field observations reveal that the equation with fractional derivative is more appropriate for describing physical phenomena than the equation which is based on the Fick's law. Thus, non-Fickian diffusion into inhomogeneous underground appears to be important in the assessment of HLW disposal. A solute transport equation with fractional derivative has been suggested and discussed in literature. However, no attempts were made to apply this equation for modeling of HLW disposal with account for the radioactive decay. In this study, we suggest the use of a novel fractional advection-diffusion equation which accounts for the effect of radioactive disintegration and for interactions between major, macro pores and fractal micro pores. This model is fundamentally different from previous proposed model of HLW, particularly in utilizing fractional derivative. Breakthrough curves numerically obtained by the present model are presented for a variety of rock types with respect to some important nuclides. Results of the calculation showed that for longer distance our model tends to be more conservative than the conventional Fickian model, therefore our model can be said to be safer.

The dissolution of quartz particles and the growth and dissolution of crystalline phases during the conversion of batch to glass potentially affects both the glass melting process and product quality. Crystals of spinel exiting the cold cap to molten glass below can be troublesome during the vitrification of iron-containing high-levelwastes. To estimate the distribution of quartz and spinel fractions within the cold cap, we used kinetic models that relate fractions of these phases to temperature and heating rate. Fitting the model equations to data showed that the heating rate, apart from affecting quartz and spinel behavior directly, also affects them indirectly via concurrent processes, such as the formation and motion of bubbles. Because of these indirect effects, it was necessary to allow one kinetic parameter (the pre-exponential factor) to vary with the heating rate. The resulting kinetic equations are sufficiently simple for the detailed modeling of batch-to-glass conversion as it occurs in glass melters. The estimated fractions and sizes of quartz and spinel particles as they leave the cold cap, determined in this study, will provide the source terms needed for modeling the behavior of these solid particles within the flow of molten glass in the melter.

This paper summarizes the work carried out to analyse different alternatives related to the inner material selection of the Spanish highlevelwaste canister for long term storage. The preliminary repository design considers granitic or clay formations, compacted bentonite sealing, corrosion allowing steel canisters and glass bead filling between the fuel assemblies and canister walls. This filling material will have the primary role of avoiding the possibility of a criticality event, which becomes an issue of major importance once the container is finally breached by corrosion and flooded by groundwater. In the first place, a complete set of requirements have been devised as evaluation criteria for candidate materials examination and selection; resulting in a compilation of demands significantly deeper and more exhaustive than any other similar work found in literature, including over 20 requirements and some other general aspects that could involve improvements in repository performance. Secondly, eight materials or material families (cast iron or steel, borosilicate glass, spinel, depleted uranium, dehydrated zeolites, hematite, phosphates and olivine) have been chosen and examined in detail, extracting some relevant conclusions. Either cast iron, borosilicate glass, spinel or depleted uranium are considered to look quite promising for the mentioned purpose. (authors)

This report presents a conceptual design for a HighLevelWaste disposal system for fuel discharged by US commercial power reactors, using the Yucca Mountain repository site recently designated by federal legislation. Principal features of the resulting conceptual design include use of unit trains for periodic removal of old spent fuel from at-reactor storage facilities, buffer storage at the repository site using dual purpose transportation/storage casks, repackaging of the spent fuel from the dual purpose transportation/storage casks directly into special-alloy disposal canisters as intact fuel assemblies, without rod consolidation, emplacement into a repository of modular design, use of excavation techniques that minimize disturbance, both mechanical and chemical, to the geologic environment, a unit rail mounted vehicle for both the transportation and emplacement of the canister from the surface facilities to the underground repository, and a cost-effectiveness computer model of Yucca Mountain and an independent cost evaluation by members of the design team. 31 refs., 58 figs., 15 tabs.

In the early 1980s, the US Department of Energy/Defense Programs (DOE/DP) initiated a project to develop a safe and efficient transportation system for defense highlevelwaste (DHLW). A long-standing objective of the DHLW transportation project is to develop a truck cask that represents the leading edge of cask technology as well as one that fully complies with all applicable DOE, Nuclear Regulatory Commission (NRC), and Department of Transportation (DOT) regulations. General Atomics (GA) designed the DHLW Truck Shipping Cask using state-of-the-art analytical techniques verified by model testing performed by Sandia National Laboratories (SNL). The analytical techniques include two approaches, inelastic analysis and elastic analysis. This topical report presents the results of the two analytical approaches and the model testing results. The purpose of this work is to show that there are two viable analytical alternatives to verify the structural adequacy of a Type B package and to obtain an NRC license. It addition, this data will help to support the future acceptance by the NRC of inelastic analysis as a tool in packaging design and licensing.

ABSTRACT This paper provides an overview of the Department of Energy's (DOE) spent nuclear fuel (SNF) and highlevelwaste (HLW) storage management. Like commercial reactor fuel, DOE's SNF and HLW were destined for the Yucca Mountain repository. In March 2010, the DOE filed a motion with the Nuclear Regulatory Commission (NRC) to withdraw the license application for the repository at Yucca Mountain. A new repository is now decades away. The default for the commercial and DOE research reactor fuel and HLW is on-site storage for the foreseeable future. Though the motion to withdraw the license application and delay opening of a repository signals extended storage, DOE's immediate plans for management of its SNF and HLW remain the same as before Yucca Mountain was designated as the repository, though it has expanded its research and development efforts to ensure safe extended storage. This paper outlines some of the proposed research that DOE is conducting and will use to enhance its storage systems and facilities.

At the West Valley Nuclear Services Company (WVNSC) in West Valley, New York, high-level radioactive waste (HLW) will be vitrified into a borosilicate glass form and poured into large, stainless steel canisters. During the filling process, volatile fission products, principally {sup 137}Cs, condense on the exterior of the canisters. The smearable contamination that remains on the canisters after they are filled and partially cooled must be removed from the canisters' exterior surfaces prior to their storage and ultimate shipment to a US Department of Energy (DOE) repository for disposal. A simple and effective method was developed for decontamination of HLW canisters. This method of chemical decontamination is applicable to a wide variety of contaminated equipment found in the nuclear industry. The process employs a reduction-oxidation system (Ce(III)/Ce(IV)) in nitric acid solution to chemically mill the surface of stainless steel, similar to the electropolishing process, but without the need for an applied electrical current. Contaminated canisters are simply immersed in the solution at controlled temperature and Ce(IV) concentration levels.

The potential repository system is intended to isolate high-level radioactive waste at Yucca Mountain according the performance objective--10 CFR 60.112. One subsystem that may contribute to achieving this objective is the sealing subsystem. This subsystem is comprised of sealing components in the shafts, ramps, underground network of drifts, and the exploratory boreholes. Sealing components can be rigid, as in the case of a shaft seal, or can be more compressible, as in the case of drift fill comprised of mined rockfill. This paper presents the preliminary seismic response of discrete sealing components in welded and nonwelded tuff. Special consideration is given to evaluating the stress in the seal, and the behavior of the interface between the seal and the rock. The seismic responses are computed using both static and dynamic analyses. Also presented is an evaluation of the maximum seismic response encountered by a drift seal with respect to the angle of incidence of the seismic wave. Mitigation strategies and seismic design considerations are proposed which can potentially enhance the overall response of the sealing component and subsequently, the performance of the overall repository system.

Sandia National Laboratories, has developed a methodology for performance assessment of deep geologic disposal of high-level nuclear waste. The applicability of this performance assessment methodology has been demonstrated for disposal in bedded salt and basalt; it has since been modified for assessment of repositories in unsaturated, fractured tuff. Changes to the methodology are primarily in the form of new or modified ground water flow and radionuclide transport codes. A new computer code, DCM3D, has been developed to model three-dimensional ground-water flow in unsaturated, fractured rock using a dual-continuum approach. The NEFTRAN 2 code has been developed to efficiently model radionuclide transport in time-dependent velocity fields, has the ability to use externally calculated pore velocities and saturations, and includes the effect of saturation dependent retardation factors. In order to use these codes together in performance-assessment-type analyses, code-coupler programs were developed to translate DCM3D output into NEFTRAN 2 input. Other portions of the performance assessment methodology were evaluated as part of modifying the methodology for tuff. The scenario methodology developed under the bedded salt program has been applied to tuff. An investigation of the applicability of uncertainty and sensitivity analysis techniques to non-linear models indicate that Monte Carlo simulation remains the most robust technique for these analyses. No changes have been recommended for the dose and health effects models, nor the biosphere transport models. 52 refs., 1 fig.

Radiological source terms for safety analyses traditionally have been curie lists of radionuclides. Converting the source term to dose values allows each radionuclide to be evaluated for its impact on dose, which is the purpose of the source term. This report is one in a series of reports establishing source terms for HighLevelWaste (HLW) by evaluating the dose impact of each radionuclide. These reports will be used in establishing the source terms to be used in HLW Safety Analysis Reports. The purpose of this report is to document the bounding element dose impact of uranium and plutonium in HLW. This technique (use of dose rather than curies) demonstrates vividly the relative importance of these nuclides in accident analyses. A large amount of available data permitted dose values to be established for uranium and plutonium; therefore, these two elements were evaluated independent of other nuclides. Solubility and adsorption data, available for these elements, allow bounding conditions to be established for their contribution to dose for various HLW processes.

Oxidative alkaline leaching has been proposed to pre-treat the high-level nuclear waste sludges to remove some of the problematic (e.g., Cr) and/or non-radioactive (e.g., Na, Al) constituents before vitrification. It is critical to understand the behavior of actinides, americium and plutonium in particular, in oxidative alkaline leaching. We have studied the leaching behavior of americium from four different sludge simulants (BiPO{sub 4}, BiPO{sub 4 modified}, Redox, PUREX) using potassium permanganate and potassium persulfate in alkaline solutions. Up to 60% of americium sorbed onto the simulants is leached from the sludges by alkaline persulfate and permanganate. The percentage of americium leached increases with [NaOH] (between 1.0 and 5.0 M). The initial rate of americium leaching by potassium persulfate increases in the order BiPO{sub 4} sludge < Redox sludge < PUREX sludge. The data are most consistent with oxidation of Am{sup 3+} in the sludge to either AmO{sub 2}{sup +} or AmO{sub 2}{sup 2+} in solution. Though neither of these species is expected to exhibit long-term stability in solution, the potential for mobilization of americium from sludge samples would have to be accommodated in the design of any oxidative leaching process for real sludge samples.

The primary goal of the US Subseabed Disposal Program (SDP) is to assess the technical and environmental feasibility of disposing of high-level nuclear waste in deep-sea sediments. The subseabed biology program is charged with assessing possible ecosystem effects of radionuclides as well as possible health effects to man from radionuclides which may be released in the deep sea and transported to the ocean surface. Current biological investigations are attempting to determine benthic community structure; benthic community metabolism; the biology of deep-sea mobile scavengers; the faunal composition of midwater nekton; rates of microbial processes, and the radiation sensitivity of deep-sea organisms. Existing models of the dispersal of radionuclides in the deep sea have not considered many of the possible biological mechanisms which may influence the movement of radionuclides. Therefore, a multi-compartment foodweb model is being developed which considers both biological and physical influences on radionuclide transport. This model will allow parametric studies to be made of the impact on the ocean environment and on man of potential releases of radionuclides.

Studies show that most Nevada residents and almost all state officials oppose the proposed high-level radioactive waste repository project at Yucca Mountain. Surveys of the public show that individual citizens view the Yucca Mountain repository as having high risk; nuclear experts, in contrast, believe the risks are very low. Policy analysts have suggested that public risk perceptions may be reduced by better program management, increased trust in the federal government, and increased economic benefits for accepting a repository. The model developed in this study is designed to examine the relationship between public perceptions of risk, trust in risk management, and potential economic impacts of the current repository program using a confirmatory multivariate method known as covariance structure analysis. The results indicate that perceptions of potential economic gains have little relationship to opposition to the repository. On the other hand, risk perceptions and the level of trust in repository management are closely related to each other and to opposition. The impacts of risk perception and trust in management on opposition to the repository result from a combination of their direct influences as well as their indirect influences operating through perceptions that the repository would have serious negative impacts on the state`s economy due to stigmatization and reduced tourism.

This report presents the results of a conceptual design of a solvent extraction process for the selective removal of {sup 137}Cs from high-level radioactive waste currently stored in underground tanks at the U.S. Department of Energy's Savannah River Site (SRS). This study establishes the need for and feasibility of deploying a simplified version of the Caustic-Side Solvent Extraction (CSSX) process; cost/benefit ratios ranging from 33 to 55 strongly support the considered deployment. Based on projected compositions, 18 million gallons of dissolved salt cake waste has been identified as having {sup 137}Cs concentrations that are substantially lower than the worst-case design basis for the CSSX system that is to be deployed as part of the Salt Waste Processing Facility (SWPF) but that does not meet the waste acceptance criteria for immobilization as grout in the Saltstone Manufacturing and Disposal Facility at SRS. Absent deployment of an alternative cesium removal process, this material will require treatment in the SWPF CSSX system, even though the cesium decontamination factor required is far less than that provided by that system. A conceptual design of a CSSX processing system designed for rapid deployment and having reduced cesium decontamination factor capability has been performed. The proposed accelerated-deployment CSSX system (CSSX-A) has been designed to have a processing rate of 3 million gallons per year, assuming 90% availability. At a more conservative availability of 75% (reflecting the novelty of the process), the annual processing capacity is 2.5 million gallons. The primary component of the process is a 20-stage cascade of centrifugal solvent extraction contactors. The decontamination and concentration factors are 40 and 15, respectively. The solvent, scrub, strip, and wash solutions are to have the same compositions as those planned for the SWPF CSSX system. As in the SWPF CSSX system, the solvent and scrub flow rates are equal. The system is

At this early stage in the conceptualization of fuel treatment and radioisotope transmutation for the disposition of nuclear wastes, it is possible to anticipate some characteristics of the waste stream resulting from the deployment of advanced technologies. Fission products and actinides cannot be completely destroyed by transmutation even with continuous purification and recycle. This is demonstrated for technetium in this analysis, but is true for all radioisotopes. Also, some of the reaction products are themselves long-lived radioactive isotopes. The purification and recycle steps produce nuclear wastes that must be planned for geologic disposal. Five radioisotopes have been identified to be produced in abundance by transmutation of technetium using fast neutrons. Four of these isotopes may be more benign than the original technetium-99 because of their longer half lives. However, one isotope, molybdenum-93 with a half life of four thousand years, may be troublesome. All of the isotopes arising from the transmutation process that end up in highlevelwaste must be examined in terms of their behavior in geologic disposal. In selecting goals for chemical separations, the technologists must consider the entire cycle of separation and transmutation before applying the performance expected in a single separation to implications concerning a repository. A separation efficiency of 0.95 can translate into the disposal of as much as 30 to 60 percent of the technetium in the repository if down stream losses are not controlled. In this case, the treatment may have little impact on anticipated off site radiation from technetium. The destruction of technetium through continuous recycle requires the cost of increased neutron dose and increased space in reactors that must be considered in design of fuel treatment systems. (authors)